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1

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project |  

Broader source: Energy.gov (indexed) [DOE]

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

2

Savannah River Site Removes Dome, Opening Reactor for Recovery Act  

Broader source: Energy.gov (indexed) [DOE]

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the workers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning More Documents & Publications Recovery Act Changes Hanford Skyline with Explosive Demolitions Recovery Act Workers Add Time Capsule Before Sealing Reactor for Hundreds

3

DOE - Office of Legacy Management -- Tatum Salt Dome Test Site...  

Office of Legacy Management (LM)

Tatum Salt Dome Test Site - MS 01 FUSRAP Considered Sites Site: Tatum Salt Dome Test Site (MS.01) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site...

4

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning  

Broader source: Energy.gov (indexed) [DOE]

American Recovery and Reinvestment American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Sa- vannah River Site this month after they safely re- moved its rusty, orange, 75-foot-tall dome. With the help of a 660-ton crane and lifting lugs, the work- ers pulled the 174,000-pound dome off the Heavy Water Components Test Reactor, capping more than 16 months of preparations. Workers will cut the dome into smaller pieces for disposal. Removal of the dome allows workers to access the 219,000-pound reactor vessel and two steam generators so they can remove and permanently dispose them onsite. Re- maining equipment will be moved to the cavity vacated by the vessel, and below-grade portions of the reactor will be

5

Advanced Test Reactor Tour  

SciTech Connect (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01T23:59:59.000Z

6

Advanced Test Reactor Tour  

ScienceCinema (OSTI)

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28T23:59:59.000Z

7

Massive Hanford Test Reactor Removed - Plutonium Recycle Test...  

Office of Environmental Management (EM)

Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed from Hanford's 300 Area Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor removed...

8

Microsoft Word - RMOTC Partners Honored for Teapot Dome Technology Test.docx  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

October 30, 2008 The Rocky Mountain Oilfield Testing Center (RMOTC) is providing the following information on local activities: RMOTC: PARTNERS HONORED FOR TEAPOT DOME TECHNOLOGY TEST Casper, Wyoming - Two partners of the Rocky Mountain Oilfield Testing Center (RMOTC) were honored at the 2008 Federal Laboratory Consortium (FLC) Mid-Continent Region meeting in Denver, Colo in September. WhisperGen LLC of New Zealand and BP America shared an Excellence in Technology Transfer award for their combined efforts in testing Stirling Cycle electrical generators for use at remote wellsites and the wide dissemination of those test results to the oil and gas industry. Stirling Cycle engines are external combustion engines which offer advantages over traditional

9

Geomechanical testing of MRIG-9 core for the potential SPR siting at the Richton salt dome.  

SciTech Connect (OSTI)

A laboratory testing program was developed to examine the mechanical behavior of salt from the Richton salt dome. The resulting information is intended for use in design and evaluation of a proposed Strategic Petroleum Reserve storage facility in that dome. Core obtained from the drill hole MRIG-9 was obtained from the Texas Bureau of Economic Geology. Mechanical properties testing included: (1) acoustic velocity wave measurements; (2) indirect tensile strength tests; (3) unconfined compressive strength tests; (4) ambient temperature quasi-static triaxial compression tests to evaluate dilational stress states at confining pressures of 725, 1450, 2175, and 2900 psi; and (5) confined triaxial creep experiments to evaluate the time-dependent behavior of the salt at axial stress differences of 4000 psi, 3500 psi, 3000 psi, 2175 psi and 2000 psi at 55 C and 4000 psi at 35 C, all at a constant confining pressure of 4000 psi. All comments, inferences, discussions of the Richton characterization and analysis are caveated by the small number of tests. Additional core and testing from a deeper well located at the proposed site is planned. The Richton rock salt is generally inhomogeneous as expressed by the density and velocity measurements with depth. In fact, we treated the salt as two populations, one clean and relatively pure (> 98% halite), the other salt with abundant (at times) anhydrite. The density has been related to the insoluble content. The limited mechanical testing completed has allowed us to conclude that the dilatational criteria are distinct for the halite-rich and other salts, and that the dilation criteria are pressure dependent. The indirect tensile strengths and unconfined compressive strengths determined are consistently lower than other coastal domal salts. The steady-state-only creep model being developed suggests that Richton salt is intermediate in creep resistance when compared to other domal and bedded salts. The results of the study provide only limited information for structural modeling needed to evaluate the integrity and safety of the proposed cavern field. This study should be augmented with more extensive testing. This report documents a series of test methods, philosophies, and empirical relationships, etc., that are used to define and extend our understanding of the mechanical behavior of the Richton salt. This understanding could be used in conjunction with planned further studies or on its own for initial assessments.

Dunn, Dennis P.; Broome, Scott Thomas; Bronowski, David R.; Bauer, Stephen J.; Hofer, John H.

2009-02-01T23:59:59.000Z

10

Savannah River Site Removes Dome, Opening Reactor for Recovery Act Decommissioning  

Broader source: Energy.gov [DOE]

American Recovery and Reinvestment Act workers achieved a significant milestone in the decommissioning of a Cold War reactor at the Savannah River Site this month after they safely removed its...

11

PIA - Advanced Test Reactor National Scientific User Facility...  

Energy Savers [EERE]

Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

12

In-situ gamma-ray site characterization of the Tatum Salt Dome Test Site in Lamar County, Mississippi  

SciTech Connect (OSTI)

Field surveys of gamma-ray emitting nuclides and soil core sampling were conducted at 12 sites on the Tatum Salt Dome Test Site and surrounding control areas to determine exposure rates from surficial radioactivity. 137Cs was the only man-made radionuclide detected and was most abundant at three off-site locations on cultivated lawns. 137Cs inventories at all of the on-site survey locations were lower than expected, given the high annual precipitation in the area. The vertical distributions were more extended than those reported for undisturbed sites. Pressurized ion chamber measurements indicated no significant differences in exposure rates on and off the test site.

Faller, S.H. (Environmental Protection Agency, Las Vegas, NV (United States))

1992-06-01T23:59:59.000Z

13

Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

14

Using reactor operating experience to improve the design of a new Broad Application Test Reactor  

SciTech Connect (OSTI)

Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

1993-07-01T23:59:59.000Z

15

Automated Test Coverage Measurement for Reactor Protection System Software  

E-Print Network [OSTI]

Automated Test Coverage Measurement for Reactor Protection System Software Implemented in Function- ing a case study using test cases prepared by domain experts for reactor protection system software) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing

16

Plutonium Recycle Test Reactor 309 B-Roll | Department of Energy  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Plutonium Recycle Test Reactor 309 B-Roll Plutonium Recycle Test Reactor 309 B-Roll Addthis Description Plutonium Recycle Test Reactor 309 B-Roll...

17

Inferences on the hydrothermal system beneath the resurgent dome in Long Valley Caldera, east-central California, USA, from recent pumping tests and geochemical sampling  

Science Journals Connector (OSTI)

Quaternary volcanic unrest has provided heat for episodic hydrothermal circulation in the Long Valley caldera, including the present-day hydrothermal system, which has been active over the past 40 kyr. The most recent period of crustal unrest in this region of east-central California began around 1980 and has included periods of intense seismicity and ground deformation. Uplift totaling more than 0.7 m has been centered on the caldera’s resurgent dome, and is best modeled by a near-vertical ellipsoidal source centered at depths of 6–7 km. Modeling of both deformation and microgravity data now suggests that (1) there are two inflation sources beneath the caldera, a shallower source 7–10 km beneath the resurgent dome and a deeper source ?15 km beneath the caldera’s south moat and (2) the shallower source may contain components of magmatic brine and gas. The Long Valley Exploration Well (LVEW), completed in 1998 on the resurgent dome, penetrates to a depth of 3 km directly above this shallower source, but bottoms in a zone of 100°C fluid with zero vertical thermal gradient. Although these results preclude extrapolations of temperatures at depths below 3 km, other information obtained from flow tests and fluid sampling at this well indicates the presence of magmatic volatiles and fault-related permeability within the metamorphic basement rocks underlying the volcanic fill. In this paper, we present recently acquired data from LVEW and compare them with information from other drill holes and thermal springs in Long Valley to delineate the likely flow paths and fluid system properties under the resurgent dome. Additional information from mineralogical assemblages in core obtained from fracture zones in LVEW documents a previous period of more vigorous and energetic fluid circulation beneath the resurgent dome. Although this system apparently died off as a result of mineral deposition and cooling (and/or deepening) of magmatic heat sources, flow testing and tidal analyses of LVEW water level data show that relatively high permeability and strain sensitivity still exist in the steeply dipping principal fracture zone penetrated at a depth of 2.6 km. The hydraulic properties of this zone would allow a pressure change induced at distances of several kilometers below the well to be observable within a matter of days. This indicates that continuous fluid pressure monitoring in the well could provide direct evidence of future intrusions of magma or high-temperature fluids at depths of 5–7 km.

Christopher D. Farrar; Michael L. Sorey; Evelyn Roeloffs; Devin L. Galloway; James F. Howle; Ronald Jacobson

2003-01-01T23:59:59.000Z

18

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect (OSTI)

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

2003-10-28T23:59:59.000Z

19

Massive Hanford Test Reactor Removed- Plutonium Recycle Test Reactor removed from Hanford’s 300 Area  

Broader source: Energy.gov [DOE]

RICHLAND, WA – Hanford’s River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy’s (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

20

Results of hydraulic tests at Gibson Dome No. 1, Elk Ridge No. 1, and E. J. Kubat boreholes, Paradox Basin, Utah  

SciTech Connect (OSTI)

Hydraulic testing was conducted in three boreholes in southeastern Utah to provide a portion of the data needed to characterize the hydrogeology of the Elk Ridge and Gibson Dome areas of the western Paradox Basin, Utah. The tests at the E. J. Kubat borehole yielded representative values of transmissivity, hydraulic conductivity, storativity, and potentiometric levels of the Leadville Limestone. Testing at Elk Ridge No. 1 provided values of similar parameters for the combined thickness of the upper Honaker Trail, Elephant Canyon, and Cedar Mesa formations. Composite transmissivities of similar zones from these borehole tests compared closely with the results of testing at borehole GD-1. A comparison of results from lab tests on core with results of extensive borehole testing at GD-1 indicates that short-term drill stem tests in a single well can provide representative estimates of bulk transmissivities and hydraulic conductivities in this field area for test zones that have a hydraulic conductivity of greater than about 1 x 10/sup -7/ cm/sec. However, lab tests produce more representative values of effective porosity and matrix permeability of individual strata. Results of lab tests and long-term borehole tests confirm that the lower Honaker Trail and upper Paradox formations have extremely low conductivities in the vicinity of the GD-1 borehole. The results of these tests were complete as of January 1981. 22 references, 29 figures, 5 tables.

Thackston, J.W.; Preslo, L.M.; Hoexter, D.E.; Donnelly, N.

1984-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

Broader source: Energy.gov (indexed) [DOE]

Advance Test Reactor Class Waiver Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

22

Engineering Test Reactor (ETR) Vessel Relocated after 50 years.  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

23

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect (OSTI)

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

24

Lava Dome | Open Energy Information  

Open Energy Info (EERE)

"coulees." Volcanic domes commonly occur within the craters or on the flanks of large composite volcanoes. The nearly circular Novarupta Dome that formed during the 1912 eruption...

25

I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a  

E-Print Network [OSTI]

I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a one­of­a­kind, tritium­fueled fusion of the Tokamak present a unique and challenging task for dismantling. Figure 1 ­ TFTR II. Project Overview The Decommissioning and Decontamination of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics

26

Internal Geology and Evolution of the Redondo Dome, Valles Caldera...  

Open Energy Info (EERE)

A detailed inventory was made of subsurface samples taken from deep geothermal test wells drilled in the resurgent Redondo dome in the Valles caldera of New Mexico. Attention...

27

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

Broader source: Energy.gov (indexed) [DOE]

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

28

TR-EDB: Test Reactor Embrittlement Data Base, Version 1  

SciTech Connect (OSTI)

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States)

1994-01-01T23:59:59.000Z

29

Detailed compositional analysis of gas seepage at the National Carbon Storage Test Site, Teapot Dome, Wyoming, USA  

Science Journals Connector (OSTI)

A baseline determination of CO2 and CH4 fluxes and soil gas concentrations of CO2 and CH4 was made over the Teapot Dome oil field in the Naval Petroleum Reserve No. 3 (NPR-3) in Wyoming, USA. This was done in anticipation of experimentation with CO2 sequestration in the Pennsylvanian Tensleep Sandstone underlying the field at a depth of 1680 m. The baseline data were collected during the winter, 2004 in order to minimize near-surface biological activity in the soil profile. The baseline data were used to select anomalous locations that may be the result of seeping thermogenic gas, along with background locations. Five 10-m holes were drilled, 3 of which had anomalous gas microseepage, and 2 were characterized as “background.” These were equipped for nested gas sampling at depths of 10-, 5-, 3-, 2-, and 1-m depths. Methane concentrations as high as 170,000 ppmv (17%) were found, along with high concentrations of C2H6, C3H8, n-C4H10, and i-C4H10. Much smaller concentrations of C2H4 and C3H6 were observed indicating the beginning of hydrocarbon oxidation in the anomalous holes. The anomalous 10-m holes also had high concentrations of isotopically enriched CO2, indicating the oxidation of hydrocarbons. Concentrations of the gases decreased upward, as expected, indicating oxidation and transport into the atmosphere. The ancient source of the gases was confirmed by 14C determinations on CO2, with radiocarbon ages approaching 38 ka within 5 m of the surface. Modeling was used to analyze the distribution of hydrocarbons in the anomalous and background 10-m holes. Diffusion alone was not sufficient to account for the hydrocarbon concentration distributions, however the data could be fit with the addition of a consumptive reaction. First-order rate constants for methanotrophic oxidation were obtained by inverse modeling. High rates of oxidation were found, particularly near the surface in the anomalous 10-m holes, demonstrating the effectiveness of the process in the attenuation of CH4 microseepage. The results also demonstrate the importance of CH4 measurements in the planning of a monitoring and verification program for geological CO2 sequestration in sites with significant remaining hydrocarbons (i.e. spent oil reservoirs).

Ronald W. Klusman

2006-01-01T23:59:59.000Z

30

E-Print Network 3.0 - application test reactor Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

test reactor Search Powered by Explorit Topic List Advanced Search Sample search results for: application test reactor Page: << < 1 2 3 4 5 > >> 1 Engineers at Western are...

31

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine  

Science Journals Connector (OSTI)

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific ...

Shijun Qin; Yuntao Song; Damao Yao; Yuanxi Wan; Songtao Wu…

2014-10-01T23:59:59.000Z

32

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Broader source: Energy.gov (indexed) [DOE]

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC)

33

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Broader source: Energy.gov (indexed) [DOE]

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

34

Tatum Dome Project, Lamar County, Mississippi  

SciTech Connect (OSTI)

The Tatum Dome site, near Hattiesburg, Mississippi, has been host to two nuclear detonations and two nonnuclear gas detonations. The top of the cavity formed by the detonations was 2660 feet below land surface and 1160 feet below the top of the salt dome. All detonations were totally contained within the salt dome in which they were fired. Activities at the site occurred between 1964 ad 1970. In 1978, NV was asked to provide proof that radiological contamination was not adversely affecting water quality in the aquifers over Tatum Dome. All except the Surficial Aquifer, which was known to be contaminated by tritium, were extensively tested to obtain hydrological data. Sufficient samples were taken to determine if contamination existed in the aquifer around the emplacement casing at surface ground zero (SGZ). Low tritium concentrations were found in the Local Aquifer, approximately 150 to 200 feet below land surface, during this project, but no radioactivity was found in any of the other aquifers above Tatum Dome. The Local Aquifer is the next aquifer below the Surficial Aquifer.

Fenske, P.R.; Humphrey, T.M. Jr.

1980-08-01T23:59:59.000Z

35

In-Situ Creep Testing Capability for the Advanced Test Reactor  

SciTech Connect (OSTI)

An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

2012-09-01T23:59:59.000Z

36

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect (OSTI)

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

37

Materials Test-2 LOCA Simulation in the NRU Reactor  

SciTech Connect (OSTI)

A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

1982-03-01T23:59:59.000Z

38

Reactor protection system with automatic self-testing and diagnostic  

DOE Patents [OSTI]

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

Gaubatz, D.C.

1996-12-17T23:59:59.000Z

39

LOCA simulation in the NRU reactor: materials test-1  

SciTech Connect (OSTI)

A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

1981-10-01T23:59:59.000Z

40

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor  

SciTech Connect (OSTI)

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

2006-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Enhanced in-pile instrumentation at the advanced test reactor  

SciTech Connect (OSTI)

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01T23:59:59.000Z

42

The Advanced Test Reactor National Scientific User Facility  

SciTech Connect (OSTI)

In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are shown in Figure 1. This article describes the ATR as well as the expanded capabilities, partnerships, and services that allow researchers to take full advantage of this national resource.

Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

43

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect (OSTI)

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

2012-03-01T23:59:59.000Z

44

E-Print Network 3.0 - aircraft shield test reactor Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

shield test reactor Search Powered by Explorit Topic List Advanced Search Sample search results for: aircraft shield test reactor Page: << < 1 2 3 4 5 > >> 1 A' Brief. History of...

45

Scaling analysis for a reactor vessel mixing test  

SciTech Connect (OSTI)

The Westinghouse AP600 advanced pressurized water reactor design uses a gravity-forced safety injection system with two nozzles in the reactor vessel downcomer. In the event of a severe overcooling transient such as a steam-line break, this system delivers boron to the core to offset positive reactivity introduced by the negative moderator defect. To determine if the system design is capable of successfully terminating this type of reactivity transient, a test of the system has been initiated. The test will utilize a 1:9 scale model of the reactor vessel and cold legs. The coolant will be modeled with air, while the safety injection fluid will be simulated with a dense gas. To determine the necessary parameters for this model, a scaling analysis was performed. The continuity, diffusion, and axial Navier-Stokes equations for the injected fluid were converted into dimensionless form. A Boussinesq formulation for turbulent viscosity was applied in these equations. This procedure identified the Richardson, mixing Reynolds, diffusion Fourier, and Euler numbers as dimensionless groups of interest. Order-of-magnitude evaluation was used to determine that the Richardson and mixing Reynolds numbers were the most significant parameters to match for a similar experiment.

Radcliff, T.D.; Parsons, J.R.; Johnson, W.S. (Univ. of Tennessee, Knoxville, TN (United States)); Ekeroth, D.E. (Westinghouse Electric Corp., Pittsburgh, PA (United States))

1993-01-01T23:59:59.000Z

46

Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor  

SciTech Connect (OSTI)

Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

1998-12-14T23:59:59.000Z

47

EPR/PTFE dosimetry for test reactor environments  

SciTech Connect (OSTI)

The use of Electron Paramagnetic Resonance (EPR) spectroscopy with materials such as alanine is well established as a technique for measurement of ionizing radiation absorbed dose in photon and electron fields such as Co-60, high-energy bremsstrahlung and electron-beam fields [1]. In fact, EPR/Alanine dosimetry has become a routine transfer standard for national standards bodies such as NIST and NPL. In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This effort consists of three stages: 1) Identification of PTFE materials that may be suitable for dosimetry applications. It was speculated that the inconsistency of EPR signatures in the earlier samples may have been due to variability in PTFE manufacturing processes. 2) Characterization of dosimetry in photon-only environments. This is necessary to establish requirements for sample preparation, operating parameters and limitations for use in well-defined and predictable environments prior to deployment in the less well-defined mixed environments of test reactors. 3) Characterization of the EPR responses obtained with PTFE in mixed neutron/photon fields. This includes evaluation of the neutron and photon contributions to response, determination of applicable of neutron fluence and photon dose ranges. This paper presents a summary of the research, a description of the EPR/PTFE dosimetry system, and recommendations for preparation and fielding of the dosimetry in photon and mixed neutron/photon environments. (authors)

Vehar, D.W.; Griffin, P.J.; Quirk, T.J. [Sandia National Laboratories, Albuquerque, NM 87185-1146 (United States)

2011-07-01T23:59:59.000Z

48

Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project  

SciTech Connect (OSTI)

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

A. B. Culp

2007-01-26T23:59:59.000Z

49

Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)  

E-Print Network [OSTI]

The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

50

Equilibrium analysis of masonry domes  

E-Print Network [OSTI]

This thesis developed a new method to analyze the structural behavior of masonry domes: the modified thrust line analysis. This graphical-based method offers several advantages to existing methods. It is the first to account ...

Lau, Wanda W

2006-01-01T23:59:59.000Z

51

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

52

Plan for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect (OSTI)

The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

Spampinato, P.T.; Walton, G.R. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Commander, J.C. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

1993-01-01T23:59:59.000Z

53

Plan for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect (OSTI)

The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D&D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D&D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D&D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D&D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

Spampinato, P.T.; Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Commander, J.C. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1993-12-31T23:59:59.000Z

54

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.  

SciTech Connect (OSTI)

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

2010-03-01T23:59:59.000Z

55

Recent results on the RIA test in IGR reactor  

SciTech Connect (OSTI)

At the 23d WRSM meeting the data base characterizing results of VVER high burnup fuel rods tests under reactivity-initiated accident (RIA) conditions was presented. Comparison of PWR and VVER failure thresholds was given also. Additional analysis of the obtained results was being carried out during 1996. The results of analysis show that the two different failure mechanisms were observed for PWR and VVER fuel rods. Some factors which can be as the possible reasons of these differences are presented. First of them is the state of preirradiated cladding. Published test data for PWR high burnup fuel rods demonstrated that the PWR high burnup fuel rods failed at the RIA test are characterized by very high level of oxidation and hydriding for the claddings. Corresponding researches were performed at Institute of Atomic Reactors (RLAR, Dimitrovgrad, Russia) for large set of VVER high burnup fuel rods. Results of these investigations show that preirradiated commercial Zr-1%Nb claddings practically keep their initial levels of oxidation and H{sub 2} concentration. Consequently the VVER preirradiated cladding must keep the high level of mechanical properties. The second reason leading to differences between failure mechanisms for two types of high burnup fuel rods can be the test conditions. Now such kind of analysis have been performed by two methods.

Asmolov, V.; Yegorova, L. [Nuclear Safety Institute, Moscow (Russian Federation)

1997-01-01T23:59:59.000Z

56

High conduction neutron absorber to simulate fast reactor environment in an existing test reactor  

SciTech Connect (OSTI)

A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

Donna Post Guillen; Larry R. Greenwood; James R. Parry

2014-10-01T23:59:59.000Z

57

Microsoft Word - 911135_0 SSC-4a Reactor Core Test Plan_rel.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for the Reactor Core Assembly Prepared by General Atomics For the Battelle Energy...

58

E-Print Network 3.0 - advanced test reactor critical facility...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test reactor critical facility Page: << < 1 2 3 4 5 > >> 1 Engineers at Western are...

59

Microsoft Word - 911136_0 SSC-4b Reactor Graphite Test Plan_rel...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Services for the Next Generation Nuclear Plant (NGNP) with Hydrogen Production Test Plan for Reactor Graphite Elements Prepared by General Atomics for the Battelle Energy...

60

A review of experiments and results from the transient reactor test (TREAT) facility.  

SciTech Connect (OSTI)

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

Deitrich, L. W.

1998-07-28T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Testing of an advanced thermochemical conversion reactor system  

SciTech Connect (OSTI)

This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

Not Available

1990-01-01T23:59:59.000Z

62

Dome Tech | Open Energy Information  

Open Energy Info (EERE)

Dome Tech Dome Tech Jump to: navigation, search Name Dome-Tech Place Edison, New Jersey Zip 8837 Sector Services Product Edison-based provider of services in engineering, energy consulting, & project development with the intention of optimising building performance. Coordinates 40.556614°, -82.866255° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":40.556614,"lon":-82.866255,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

63

Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)  

SciTech Connect (OSTI)

This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

Bradley K. Heath

2014-03-01T23:59:59.000Z

64

Salt dome discoveries mounting in Mississippi  

SciTech Connect (OSTI)

Exploratory drilling around piercement salt domes in Mississippi has met with a string of successes in recent months. Exploration of these salt features is reported to have been initiated through the review of non-proprietary, 2D seismic data and subsurface control. This preliminary data and work were then selectively upgraded by the acquisition of additional, generally higher quality, conventional 2D seismic lines. This current flurry of successful exploration and ensuing development drilling by Amerada Hess Corp. on the flanks of salt domes in Mississippi has resulted in a number of significant Hosston discoveries/producers at: Carson salt dome in Jefferson Davis County; Dry Creek salt dome in Covington County, Midway salt dome in lamar County, Monticello salt dome in Lawrence County, and Prentiss salt dome in Jefferson Davis County. The resulting production from these fields is gas and condensate, with wells being completed on 640 acre production units.

Ericksen, R.L. [Mississippi Office of Geology, Jackson, MS (United States)

1996-06-17T23:59:59.000Z

65

E-Print Network 3.0 - advanced test idaho reactor Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Powered by Explorit Topic List Advanced Search Sample search results for: advanced test idaho reactor Page: << < 1 2 3 4 5 > >> 1 Point of Contact: Doug Kothe CASL Director...

66

E-Print Network 3.0 - advanced test reactor Sample Search Results  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

20 03012006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http:english.people.com.cn200603...

67

Preliminary Cost Assessment and Compare of China Fusion Engineering Test Reactor  

Science Journals Connector (OSTI)

Full superconducting tokamak and water-cooling Cu magnets tokamak are two options proposed for China Fusion Engineering Test Reactor (CFETR). Based on the concept design ... Program for Parameters Optimization an...

Dehong Chen; Jieqiong Jiang; Yawei Hou; Wenxue Duan; Muyi Ni…

2014-09-01T23:59:59.000Z

68

Dome-Tech and Merck Teaming Profile  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Dome-Tech, Inc. Merck & Co., Inc. Dome-Tech, Inc. Merck & Co., Inc. 510 Thornall Street One Merck Drive Edison, New Jersey 08837 Whitehouse Station, New Jersey 08889 Business: Energy Engineering Consulting Business: Pharmaceutical Manufacturing Ed Liberty Rob Colucci Vice President, Energy Advisors Senior Director, Energy and Sustainability Phone: 732-590-0122 ext. 165 Phone: 908-423-4971 Email: e_liberty@dome-tech.com Email: robert_colucci@merck.com Dome-Tech installs 1.6 MW DC solar power tracking system at Merck to reduce CO 2 emissions Project Scope Dome-Tech installed a 1.6 megawatt solar photovoltaic (PV) system at Merck's corporate headquarters in New Jersey - with no capital investment by Merck. Project Summary Dome-Tech worked with SunPower to install 7,000 solar panels on 7 acres of land at Merck's

69

Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)  

SciTech Connect (OSTI)

A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

2012-07-01T23:59:59.000Z

70

Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion  

Science Journals Connector (OSTI)

The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system extensive testing of the various proposed concepts will be required. In today’s environmental safety and health culture full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA’s schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

Charles S. Olsen; Henry Welland; James Abraschoff; Kenneth Thoms

1993-01-01T23:59:59.000Z

71

Resurgent Dome Complex | Open Energy Information  

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Resurgent Dome Complex Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Print PDF Resurgent Dome Complex Dictionary.png Resurgent Dome Complex: Resurgent domes are encountered near the center of many caldera depressions, and form via uplift of the caldera valley floor due to movement in the underlying magma chamber. Resurgent domes typically host numerous deformation structures that act as conduits for hydrothermal fluids in the shallow crust. Other definitions:Wikipedia Reegle Topographic Features List of topographic features commonly encountered in geothermal resource

72

Light Water Reactor Fuel Cladding Research and Testing | ornl.gov  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

73

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information  

SciTech Connect (OSTI)

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

74

Initial Confinement Studies of Ohmically Heated Plasmas in the Tokamak Fusion Test Reactor  

Science Journals Connector (OSTI)

Initial operation of the tokamak fusion test reactor has concentrated upon confinement studies of Ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (?E) are 0.1-0.2 s for a line-average density range (n¯e) of (1-2.5)×1019 m-3 with electron temperatures of Te(0)?1.2-2.2 keV, ion temperatures of Ti(0)?0.9-1.5 keV, and Zeff?3. A comparison of Princeton large torus, poloidal divertor experiment, and tokamak fusion test reactor plasma confinement supports a dimension-cubed scaling law.

P. C. Efthimion et al.

1984-04-23T23:59:59.000Z

75

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

SciTech Connect (OSTI)

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

76

Continuous-flow stirred-tank reactor 20-L demonstration test: Final report  

SciTech Connect (OSTI)

One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

Lee, D.D.; Collins, J.L.

2000-02-01T23:59:59.000Z

77

Energy Department Sells Historic Teapot Dome Oilfield  

Broader source: Energy.gov [DOE]

Today, the Energy Department finalized the sale of the historic Teapot Dome Oilfield located 35 miles north of Casper, Wyoming to Stranded Oil Resources Corporation, a subsidiary of Allegheny Capital Corporation.

78

2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

Mike Lewis

2014-02-01T23:59:59.000Z

79

Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect (OSTI)

Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

1984-06-01T23:59:59.000Z

80

Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine  

SciTech Connect (OSTI)

This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

Reilly, Raymond W.

2012-07-30T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

SciTech Connect (OSTI)

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

2009-09-15T23:59:59.000Z

82

Final report on decommissioning boreholes and wellsite restoration, Gulf Coast Interior Salt Domes of Mississippi  

SciTech Connect (OSTI)

In 1978, eight salt domes in Texas, Louisiana, and Mississippi were identified for study as potential locations for a nuclear waste repository as part of the National Waste Terminal Storage (NWTS) program. Three domes were selected in Mississippi for ``area characterization`` phase study as follows: Lampton Dome near Columbia, Cypress Creek Dome near New Augusta, and Richton Dome near Richton. The purpose of the studies was to acquire geologic and geohydrologic information from shallow and deep drilling investigations to enable selection of sites suitable for more intensive study. Eleven deep well sites were selected for multiple-well installations to acquire information on the lithologic and hydraulic properties of regional aquifers. In 1986, the Gulf Coast salt domes were eliminated from further consideration for repository development by the selection of three candidate sites in other regions of the country. In 1987, well plugging and restoration of these deferred sites became a closeout activity. The primary objectives of this activity are to plug and abandon all wells and boreholes in accordance with state regulations, restore all drilling sites to as near original condition as feasible, and convey to landowners any wells on their property that they choose to maintain. This report describes the activities undertaken to accomplish these objectives, as outlines in Activity Plan 1--2, ``Activity Plan for Well Plugging and Site Restoration of Test Hole Sites in Mississippi.``

Not Available

1989-04-01T23:59:59.000Z

83

RERTR Program: goals, progress and plans. [Reduced Enrichment Research and Test Reactor  

SciTech Connect (OSTI)

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the nearly null value of 1982 to the 7.0 g U/cm/sup 3/ which will be reached in early 1989. The technical needs of research reactors for HEU exports are also estimated to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1984-09-25T23:59:59.000Z

84

Results of the DF-4 BWR (boiling water reactor) control blade-channel box test  

SciTech Connect (OSTI)

The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

Gauntt, R.O.; Gasser, R.D.

1990-10-01T23:59:59.000Z

85

Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors  

SciTech Connect (OSTI)

The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

Not Available

1993-07-01T23:59:59.000Z

86

Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors  

SciTech Connect (OSTI)

Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

2012-02-01T23:59:59.000Z

87

Geology of Damon Mound Salt Dome, Texas  

SciTech Connect (OSTI)

Geological investigation of the stratigraphy, cap-rock characteristics, deformation and growth history, and growth rate of a shallow coastal diapir. Damon Mound salt dome, located in Brazoria County, has salt less than 600 feet and cap rock less than 100 feet below the surface; a quarry over the dome provides excellent exposures of cap rock as well as overlying Oligocene to Pleistocene strata. These conditions make it ideal as a case study for other coastal diapirs that lack bedrock exposures. Such investigations are important because salt domes are currently being considered by chemical waste disposal companies as possible storage and disposal sites. In this book, the author reviews previous research, presents additional data on the subsurface and surface geology at Damon Mound, and evaluates Oligocene to post-Pleistocene diapir growth.

Collins, E.W.

1989-01-01T23:59:59.000Z

88

Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report  

SciTech Connect (OSTI)

Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

1982-03-01T23:59:59.000Z

89

Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study  

SciTech Connect (OSTI)

Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

2012-08-01T23:59:59.000Z

90

Acoustic Emission Monitoring of ASME Section III Hydrostatic Test: Watts Bar Unit 1 Nuclear Reactor  

SciTech Connect (OSTI)

Through the cooperation of the Tennessee Valley Authority, Pacific Northwest Laboratory has installed instrumentation on Watts Bar Nuclear Power Plant Unit 1 for the purpose of test and evaluation of acoustic emission (AE) monitoring of nuclear reactor pressure vessels and piping for flaw detection. This report describes the acoustic emission monitoring performed during the ASME Section III hydrostatic testing of Watts Bar Nuclear Power Plant Unit 1 and the results obtained. Highlights of the results are: • Spontaneous AE was detected from a nozzle area during final pressurization. • Evaluation of the apparent source of the spontaneous AE using an empirically derived AE/fracture mechanics relationship agreed within a factor of two with an evaluation by ASME Section XI Code procedures. • AE was detected from a fracture specimen which was pressure coupled to the 10-inch accumulator nozzle. This provided reassurance of adequate system sensitivity. • High background noise was observed when all four reactor coolant pumps were operating. Work is continuing at Watts Bar Unit 1 toward AE monitoring hot functional testing and subsequently monitoring during reactor operation.

Hutton,, P. H.; Taylor,, T. T.; Dawson,, J. F.; Pappas,, R. A.; Kurtz,, R. J.

1982-06-01T23:59:59.000Z

91

Radar investigation of the Hockley salt dome  

E-Print Network [OSTI]

: Geophysics RADAR INVESTIGATION OF THE HOCKLEY SALT DOME A Thesis by UAMES ANDREW HLUCHANEK A'pproved as to style and content by: (Head of Departme t ? Member) May 1. 973 ABSTRACT Radar investigation of the Hockley Salt Dome. . (Nay, 1973) James... Andrew Hluchanek, B. S. , Texas A&M University Directed by: Dr. Robert R. Unterberger Radar probing through salt was accomplished at 17 radar stations established in the United Salt Company mine at Hockley, Texas. The top of the salt dom is mapped...

Hluchanek, James Andrew

2012-06-07T23:59:59.000Z

92

Axial effects of xenon-samarium poisoning in the advanced test reactor  

SciTech Connect (OSTI)

The paper details an analytical study of the time-dependent behavior in the spatial distributions of xenon and samarium fission product poisons in the Advanced Test Reactor (ATR) during operation and after shutdown. The results of this study provide insight into the behavior and significance of the changing spatial distributions of fission product poisons with respect to the prediction of shim positions at critical for reactor restart after a xenon shutdown. The study was performed with the PDQ neutron diffusion theory code and ENDF/B-V cross sections using a one-dimensional radial model of an ATR lobe and a two-dimensional radial-axial (RZ) model of an ATR lobe. The PDQ results were supported by a review of the basic differential equations, which describe the buildup and decay of the xenon and samarium fission product poisons and precursors. The ATR is a 250-MW, uranium-aluminum-fueled reactor used to study the effects of irradiation on reactor materials. Forty highly enriched uranium fuel elements are arranged in a serpentine configuration within the compact core resulting in a very high power density of (1.0 MW/[ell] of core).

Auslander, D.J.; Smith, A.C.; McCracken, R.T. (Idaho National Engineering Lab., Idaho Falls (United States))

1990-01-01T23:59:59.000Z

93

Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long  

Open Energy Info (EERE)

Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long Valley Caldera, East-Central California, Usa, From Recent Pumping Tests And Geochemical Sampling Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Inferences On The Hydrothermal System Beneath The Resurgent Dome In Long Valley Caldera, East-Central California, Usa, From Recent Pumping Tests And Geochemical Sampling Details Activities (6) Areas (1) Regions (0) Abstract: Quaternary volcanic unrest has provided heat for episodic hydrothermal circulation in the Long Valley caldera, including the present-day hydrothermal system, which has been active over the past 40 kyr. The most recent period of crustal unrest in this region of east-central California began around 1980 and has included periods of

94

Testing of Passive Safety System Performance for Higher Power Advanced Reactors  

SciTech Connect (OSTI)

This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

2004-12-31T23:59:59.000Z

95

An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor  

SciTech Connect (OSTI)

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

Yoder Jr, Graydon L [ORNL] [ORNL; Aaron, Adam M [ORNL] [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Kisner, Roger A [ORNL] [ORNL; Peretz, Fred J [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Wilgen, John B [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL

2014-01-01T23:59:59.000Z

96

An Approach to Mapping of Shallow Petroleum Reservoirs Using Integrated Conventional 3D and Shallow P- and SH-Wave Seismic Reflection Methods at Teapot Dome Field in Casper, Wyoming.  

E-Print Network [OSTI]

??Using the famous Teapot Dome oil field in Casper, Wyoming, USA as a test case, we demonstrate how high-resolution compressional (P) and horizontally polarized shear… (more)

Okojie-Ayoro, Anita Onohuome 1981-

2007-01-01T23:59:59.000Z

97

Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing  

Science Journals Connector (OSTI)

Abstract Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K , followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

E. Korobkina; G. Medlin; B. Wehring; A.I. Hawari; P.R. Huffman; A.R. Young; B. Beaumont; G. Palmquist

2014-01-01T23:59:59.000Z

98

The RERTR (Reduced Enrichment Research and Test Reactor) program: A progress report  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U/sub 3/Si/sub 2/-Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product /sup 99/Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program.

Travelli, A.

1986-11-01T23:59:59.000Z

99

Occurrence of gypsum in Gulf coast salt domes  

Science Journals Connector (OSTI)

...Occurrence of gypsum in Gulf coast salt domes Barton Donald Clinton...OF GYPSUM IN THE GULF COAST SALT DOMES. Sir: On accountof thepaucityof...concerningtheoccurrenceof gypsumandanhydriteon the salt domes. The followingremarksmay...5o-footsill of saltat 3,350feetat Palangana,possiblyalsoStrattonRidge...

Donald Clinton Barton

100

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

Science Journals Connector (OSTI)

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the...

S. I. Gusev; V. N. Karpov; A. N. Kiselev; V. I. Kochkin

2009-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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101

Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor  

SciTech Connect (OSTI)

The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (authors)

Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia [Department of Nuclear Engineering, Politecnico di Milano, Via Ponzio, 34/3, 20133 Milano (Italy)

2004-07-01T23:59:59.000Z

102

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting  

SciTech Connect (OSTI)

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

NONE

2008-07-15T23:59:59.000Z

103

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop  

SciTech Connect (OSTI)

This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

Donna Post Guillen

2012-11-01T23:59:59.000Z

104

Progress of the RERTR (Reduced Enrichment Research and Test Reactor) Program in 1989  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs.

Travelli, A.

1989-01-01T23:59:59.000Z

105

Methods for nondestructive testing of austenitic high-temperature gas-cooled reactor components  

SciTech Connect (OSTI)

Safety-relevant components of high-temperature gas-cooled reactor components are mostly fabricated in nickel-based alloys and austenitic materials like Inconel-617, Hastelloy-X, Nimonic-86, or Incoloy-800H. Compared to ferritic steels, these austenitic materials can have a coarse-grained microstructure, especially in weldments and castings. Coarse-grained or elastic anisotropic materials are difficult to inspect with ultrasonics due to strong attenuation, high noise level (scattering, ''grass'' indications), and sound beam distortions (skewing, splitting, and mode conversion). Only few results dealing with the nondestructive testing of nickel-based alloys are known. The problem area, solutions, and first experiences are reported.

Gobbels, K.; Kapitza, H.

1984-09-01T23:59:59.000Z

106

E-Print Network 3.0 - aerospace system test reactor Sample Search...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Engineering 15 www.eprg.group.cam.ac.uk EPRGWORKINGPAPER Summary: Accelerator-Driven Subcritical Reactor Park Michel-Alexandre Cardin1 Engineering Systems Division... and reactor...

107

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research  

SciTech Connect (OSTI)

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

108

Ge atom distribution in buried dome islands  

SciTech Connect (OSTI)

Laser-assisted atom probe tomography microscopy is used to provide direct and quantitative compositional measurements of tri-dimensional Ge distribution in Ge dome islands buried by Si. Sub-nanometer spatial resolution 3D imaging shows that islands keep their facets after deposition of the Si cap, and that the island/substrate/Si cap interfaces are abrupt. The core of the domes contains 55% of Ge, while the island shell exhibits a constant composition of 15% of Ge. The {l_brace}113{r_brace} facets of the islands present a Ge enrichment up to 35%. The wetting layer composition is not homogeneous, varying from 9.5% to 30% of Ge.

Portavoce, A.; Berbezier, I.; Ronda, A.; Mangelinck, D. [CNRS, IM2NP, Case 142, 13397 Marseille Cedex 20 (France); Hoummada, K. [Aix-Marseille Universite, IM2NP, Case 142, 13397 Marseille Cedex 20 (France)

2012-04-16T23:59:59.000Z

109

The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

Travelli, A.

1987-01-01T23:59:59.000Z

110

Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect (OSTI)

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

Travelli, A.

1988-01-01T23:59:59.000Z

111

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, Terry Alan

2002-06-01T23:59:59.000Z

112

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

Tomberlin, T.A.

2002-06-19T23:59:59.000Z

113

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations  

SciTech Connect (OSTI)

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

2014-01-01T23:59:59.000Z

114

Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors  

SciTech Connect (OSTI)

The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO(subscript 2)-fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO (subscript 2)-fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The test measured a peak power output of 530 watts(e) or 7.1 watts/cm (superscript 2) at an efficiency of 11.5%. There are three copies in the file. Cross-Reference a copy FSC-ESD-217-94-529 in the ESD files with a CID #8574.

Schock, Alfred

1994-06-01T23:59:59.000Z

115

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

SciTech Connect (OSTI)

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

116

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask  

SciTech Connect (OSTI)

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Romano, T.

1997-09-29T23:59:59.000Z

117

Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington  

SciTech Connect (OSTI)

During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

S.J. Roberts

2007-03-20T23:59:59.000Z

118

Information Sheet 2011/1/29 DOT Folddown Protective Domes  

E-Print Network [OSTI]

to protect solar telescopes by providing a dome that can be folded down flat to minimize temperature and wind effects, this concept can also be used for other situations. The two solar telescope domes that were built are located on top of towers at an altitude of approximately 2300 m on mountains

Rutten, Rob

119

Status of the RERTR program: overview, progress and plans. [Reduced Enrighment Research and Test Reactor  

SciTech Connect (OSTI)

The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a summary of the accomplishments which the RERTR Program had achieved by the end of 1984 with its many international partners, emphasis is placed on the progress achieved during 1985 and on current plans and schedules. A new miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was fabricated and is well into irradiation. The whole-core ORR demonstration is scheduled to begin in November 1985, with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/. Altogether, 921 full-size test and prototype elements have been ordered for fabrication with reduced enrichment and the new technologies. Qualification of U/sub 3/Si-Al fuel with approx.7 g U/cm/sup 3/ is still projected for 1989. This progress could not have been achieved without the close international cooperation which has existed since the beginning, and whose continuation and intensification will be essential to the achievement of the long-term RERTR goals.

Travelli, A.

1985-01-01T23:59:59.000Z

120

Quasi-static rock mechanics data for rocksalt from three Strategic Petroleum Reserve domes  

SciTech Connect (OSTI)

Triaxial compression and extension experiments have been run on rocksalt samples from three Strategic Petroleum Reserve (SPR) domes. Seventeen quasi-static tests were loaded at mean stress rates of .66 to 1.04 psi/sec (4.5 to 7.2 kPa/sec), confining pressures of 14.5 to 2000 psi (0.1 to 13.8 MPa) and temperatures of 22 to 100/sup 0/C. Eleven of the test specimens were from Bryan Mound, Texas, and three each were from Bayou Choctaw, Louisiana, and West Hackberry, Louisiana. In general, the resulting mechanical data from the three domes are similar, and they are consistent with previously published data. Ultimate sample strengths are directly related to confining pressure (least principal stress) and indirectly related to temperature, while ductility increases with both pressure and temperature.

Price, R.H.; Wawersik, W.R.; Hannum, D.W.; Zirzow, J.A.

1981-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Baseline risk assessment of the perched water system at the INEL test reactor area  

SciTech Connect (OSTI)

A baseline health risk assessment (HRA) was prepared to evaluate potential risks to human health and the environment posed by the Perched Water System (PWS) at the Test Reactor Area (TRA). The PWS has been designated Operable Unit 2-12, one of the 13 operable units identified at TRA. During the period from 1962 to 1990, a total of 6770 million gal of water were discharged from the TRA to unlined surface ponds. Wastewater discharged to the surface ponds at TRA percolates downward through the surficial alluvium and the underlying basalt bedrock. A resulting shallow perched water zone has formed at the interface between the surficial sediments and the underlying basalt. Further downward movement of groundwater is again impeded by a low-permeability layer of silt, clay, and sand encountered at a depth of [approximately]150 ft. The deep perched water zone occurs on top of this low-permeability interbed. An evaluation was made as to whether potential risks for the PWS could justify implementing a remedial action. The risk evaluation consisted of two parts, the human health evaluation and the ecological evaluation.

Gordon, J.W.; Sinton, P.O. (Dames Moore, Denver, CO (United States)); Jensen, N. (DOE, Idaho Falls, ID (United States)); McCormick, S. (Idaho National Engineering Lab., Idaho Falls, ID (United States))

1993-01-01T23:59:59.000Z

122

Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project  

SciTech Connect (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

123

Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project  

SciTech Connect (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

124

Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project  

SciTech Connect (OSTI)

In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

Noel Duckwitz

2011-06-01T23:59:59.000Z

125

Geology of Bravo Dome carbon dioxide gas field, New Mexico  

SciTech Connect (OSTI)

The Bravo Dome carbon dioxide gas field is located in Union and Harding Counties of northeast New Mexico. The Bravo Dome field covers approximately 800,000 acres, but areal boundaries of the field have not been fully defined. Production in 1989 was 113 bcf of gas from 272 wells. Cumulative production at the end of 1989 was 626 bcf. Estimated recoverable reserves are more than 10 tcf. The gas is 98-99% CO{sub 2}. Most CO{sub 2} produced from Bravo Dome is used for enhanced oil recovery in the Permian basin. The Bravo Dome is a faulted, southeast-plunging, basement-cored anticlinal nose. It is bordered on the east and south by large high-angle faults of Pennsylvanian and Wolfcampian (Early Permian) age. The principal reservoir in the Bravo Dome field is the Tubb sandstone (Leonardian-Permian) at depths of 1,900 to 2,950 ft. The Tubb consists of 0-400 ft of fine- to medium-grained, well-sorted, orange feldspathic sandstone. It rests unconformably on Precambrian basement on the highest parts of the Bravo Dome and is not offset by late Paleozoic faults that form the dome. The Cimmaron Anhydrite (Leonardian-Permian) conformably overlies the Tubb and is a vertical seal. The trap at Bravo Dome has structural and stratigraphic aspects. Drape of Tubb sandstone over the dome created structural closure on the northeast, southeast, and southwest flanks of the field. Trapping on the northwest flank of the field is associated with regional northwest thinning of the Tubb.

Broadhead, R.F. (New Mexico Bureau of Mines and Mineral Resources, Socorro (United States))

1991-03-01T23:59:59.000Z

126

The technique and preliminary results of LEU U-Mo full-size IRT type fuel testing in the MIR reactor  

SciTech Connect (OSTI)

In March 2007 in-pile testing of LEU U-Mo full-size IRT type fuel elements was started in the MIR reactor. Four prototype fuel elements for Uzbekistan WWR SM reactor are being tested simultaneously - two of tube type design and two of pin type design. The dismountable irradiation devices were constructed for intermediate reloading and inspection of fuel elements during reactor testing. The objective of the test is to obtain the experimental results for determination of more reliable design and licensing fuel elements for conversion of the WWR SM reactor. The heat power of fuel elements is measured on-line by thermal balance method. The distribution of fission density and burn-up of uranium in the volume of elements are calculated by using the MIR reactor MCU code (Monte-Carlo) model. In this paper the design of fuel elements, the technique, main parameters and preliminary results are described. (author)

Izhutov, A.L.; Starkov, V.A.; Pimenov, V.V.; Fedoseev, V.Ye. [Research Reactor Complex, RIAR, 433510, Dimitrovgrad-10, Ulyanovsk Region (Russian Federation); Dobrikova, I.V.; Vatulin, A.V.; Suprun, V.B. [A.A. Bochvar All-Russian Scientific Research Institute of Inorganic Materials, P. O. Box 369, 123060, Moscow (Russian Federation); Kartashov, Ye.F.; Lukichev, V.A. [Research and Development Institute of Nuclear Energy and Industry, P. O. Box 788, 107014, Moscow (Russian Federation); Troyanov, V.M.; Enin, A.A.; Tkachev, A.A. [OAO 'TVEL' 119017, ul. B. Ordinka 24/26, Moscow (Russian Federation)

2008-07-15T23:59:59.000Z

127

Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors  

SciTech Connect (OSTI)

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 hours. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. Design of a capsule using boron carbide enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to {sup 235}U fission.

Greenwood, Lawrence R.; Wittman, Richard S.; Pierson, Bruce D.; Metz, Lori A.; Payne, Rosara F.; Finn, Erin C.; Friese, Judah I.

2012-03-01T23:59:59.000Z

128

Final report on decommissioning of wells, boreholes, and tiltmeter sites, Gulf Coast Interior Salt Domes of Louisiana  

SciTech Connect (OSTI)

In the late 1970s, test holes were drilled in northern Louisiana in the vicinity of Vacherie and Rayburn`s Salt Domes as part of the Department of Energy`s (DOE) National Waste Terminal Storage (NWTS) (rename the Civilian Radioactive Waste Management (CRWM)) program. The purpose of the program was to evaluate the suitability of salt domes for long term storage or disposal of high-level nuclear waste. The Institute for Environmental Studies at Louisiana State University (IES/LSU) and Law Engineering Testing Company (LETCo) of Marietta, Georgia performed the initial field studies. In 1982, DOE awarded a contract to the Earth Technology Corporation (TETC) of Long Beach, California to continue the Gulf Coast Salt Dome studies. In 1986, DOE deferred salt domes from further consideration as repository sites. This report describes test well plugging and site abandonment activities performed by SWEC in accordance with Activity Plan (AP) 1--3, Well Plugging and Site Restoration of Work Sites in Louisiana. The objective of the work outlined in this AP was to return test sites to as near original condition as possible by plugging boreholes, removing equipment, regrading, and seeding. Appendices to this report contain forms required by State of Louisiana, used by SWEC to document decommissioning activities, and pertinent documentation related to lease/access agreements.

Not Available

1989-07-01T23:59:59.000Z

129

Basic Test of A 3-Phase Superconducting Fault Current Limiting Reactor  

Science Journals Connector (OSTI)

A novel 3-phase superconducting fault current limiting reactor (SCFCLR) for power system ... some experiments, the fundamental behavior of the limiter is confirmed. In the experiment, two ... exhibits very small ...

H. Kado; T. Ishigohka

1990-01-01T23:59:59.000Z

130

Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests  

E-Print Network [OSTI]

This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

International Electrotechnical Commission. Geneva

2005-01-01T23:59:59.000Z

131

Distribution of quaternary rhyolite dome of the Coso Range, California:  

Open Energy Info (EERE)

of quaternary rhyolite dome of the Coso Range, California: of quaternary rhyolite dome of the Coso Range, California: Implications for extent of the geothermal anomaly Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Distribution of quaternary rhyolite dome of the Coso Range, California: Implications for extent of the geothermal anomaly Details Activities (1) Areas (1) Regions (0) Abstract: Thirty-eight separate domes and flows of phenocryst-poor, high-silica rhyolite of similar major element chemical composition were erupted over the past 1 m.y. from vents arranged in a crudely S-shaped array atop a granitic horst in the Coso Range, California. Most of the extrusions are probably less than about 0.3 m.y. old. The area is one of Quaternary basaltic volcanism and crustal extension. The central part of

132

Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF  

SciTech Connect (OSTI)

The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could actually help extend the life of a fuel pin in a “long pin” reactor design to a higher peak fuel burnup.

D. L. Porter; H.C. Tsai

2012-08-01T23:59:59.000Z

133

Icosadeltahedral geometry of fullerenes, viruses and geodesic domes  

E-Print Network [OSTI]

I discuss the symmetry of fullerenes, viruses and geodesic domes within a unified framework of icosadeltahedral representation of these objects. The icosadeltahedral symmetry is explained in details by examination of all of these structures. Using Euler's theorem on polyhedra, it is shown how to calculate the number of vertices, edges, and faces in domes, and number of atoms, bonds and pentagonal and hexagonal rings in fullerenes. Caspar-Klug classification of viruses is elaborated as a specific case of icosadeltahedral geometry.

Antonio Siber

2007-11-22T23:59:59.000Z

134

Fission product transport and behavior during two postulated loss-of-flow transients in the Advanced Test Reactor  

SciTech Connect (OSTI)

The fission product behavior during two postulated loss-of-flow accidents (leading to high- and low-pressure core degradations) in the Advanced Test Reactor (ATR) has been analyzed. These transients are designated ATR transients LCP 15 (high pressure) and LPP9 (low pressure). Normally, transients of this nature would be easily mitigated using existing safety systems and procedures. In these analyses, failure of these safety systems was assumed so that core degradation and fission product release could be studied. A probabilistic risk analysis was performed that indicated that the probability of occurrence for these two transients is on the order of 10[sup [minus]5] and 10[sup [minus]7] per reactor year for LCP15 and LPP9, respectively. The fission product behavior analysis included calculations of the gaseous and highly volatile fission product (xenon, krypton, cesium, iodine, and tellurium) inventories in the fuel before accident initiation, release of the fission products from the fuel into the reactor vessel during core melt, the probable chemical forms, and transport of the fission products from the core through the reactor vessel and existing piping to the confinement. In addition to a base-case analysis of fission product behavior, a series of analyses was performed to determine the sensitivity of fission product release to several parameters including steam flow rate, (structural) aluminum oxidation, and initial aerosol size. The base-case analyses indicate that the volatile fission products (excluding the noble gases) will be transported as condensed species on zinc aerosols.

Adams, J.P.; Carboneau, M.L.; Hagrman, D.L. (Idaho National Engineering Lab. EG and G Idaho, Idaho Falls, ID (United States))

1993-07-01T23:59:59.000Z

135

Scoping analysis of the Advanced Test Reactor using SN2ND  

SciTech Connect (OSTI)

A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR. The successful use of such codes in a coupled dynamics simulation can impact what experiments are performed and what power levels are permitted during those experiments at the ATR. The advantages of the SN2ND solver over comparable neutronics tools are its superior parallel performance and demonstrated accuracy on large scale homogeneous and heterogeneous reactor geometries. However, it should be noted that virtually no effort from this project was spent constructing a proper cross section generation methodology for the ATR usable in the SN2ND solver. While attempts were made to use cross section data derived from SCALE, the minimal number of compositional cross section sets were generated to be consistent with the reference Monte Carlo input specification. The accuracy of any deterministic transport solver is impacted by such an approach and clearly it causes substantial errors in this work. The reasoning behind this decision is justified given the overall funding dedicated to the task (two months) and the real focus of the work: can modern deterministic tools actually treat complex facilities like the ATR with heterogeneous geometry modeling. SN2ND has been demonstrated to solve problems with upwards of one trillion degrees of freedom which translates to tens of millions of finite elements, hundreds of angles, and hundreds of energy groups, resulting in a very high-fidelity model of the system unachievable by most deterministic transport codes today. A space-angle convergence study was conducted to determine the meshing and angular cubature requirements for the ATR, and also to demonstrate the feasibility of performing this analysis with a deterministic transport code capable of modeling heterogeneous geometries. The work performed indicates that a minimum of 260,000 linear finite elements combined with a L3T11 cubature (96 angles on the sphere) is required for both eigenvalue and flux convergence of the ATR. A critical finding was that the fuel meat and water channels must each be meshed with at least 3 'radial zones' for accurate flux convergence. A small number of 3D calculations were also performed to show axial mesh and eigenvalue convergence for a full core problem. Finally, a brief analysis was performed with different cross sections sets generated from DRAGON and SCALE, and the findings show that more effort will be required to improve the multigroup cross section generation process. The total number of degrees of freedom for a converged 27 group, 2D ATR problem is {approx}340 million. This number increases to {approx}25 billion for a 3D ATR problem. This scoping study shows that both 2D and 3D calculations are well within the capabilities of the current SN2ND solver, given the availability of a large-scale computing center such as BlueGene/P. However, dynamics calculations are not realistic without the implementation of improvements in the solver.

Wolters, E.; Smith, M. (NE NEAMS PROGRAM); ( SC)

2012-07-26T23:59:59.000Z

136

Cold domes over the warm pool: a study of the properties of cold domes produced by mesoscale convective systems during TOGA COARE  

E-Print Network [OSTI]

in specific humidity within their cold domes. As a result, increases in the latent heat fluxes observed in the cold domes were more dependent on the increases in wind speed. Stronger cold domes (cool, dry and strong winds) were associated with squall line MCSs...

Caesar, Kathy-Ann Lois

2012-06-07T23:59:59.000Z

137

Former Reactor Facilities Surveillance and Maintenance and  

E-Print Network [OSTI]

and Dark (2010) Majority of remaining radiation contained in reactor vessel (inside biological shield window damage Minor roof leaks to former office areas Stack and Grounds: Stack drain tank Safety of confinement dome Replaced exhaust fan Security system hardware/software upgrade BGRR: Covered vents above 2

Ohta, Shigemi

138

MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT  

SciTech Connect (OSTI)

The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

Vinson, D.

2010-07-11T23:59:59.000Z

139

Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory  

SciTech Connect (OSTI)

The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

2007-10-01T23:59:59.000Z

140

Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management  

SciTech Connect (OSTI)

The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety basis. The need for a design basis reconstitution program for the ATR has been identified along with the use of sound configuration management principles in order to support safe and efficient facility operation.

G. L. Sharp; R. T. McCracken

2004-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
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141

10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement  

SciTech Connect (OSTI)

The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

David E. Korns

2012-05-01T23:59:59.000Z

142

New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor  

Science Journals Connector (OSTI)

The original description of alpha particle driven instabilities in the Tokamak Fusion Test Reactor in terms of toroidal Alfvén eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the antiballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time.

R. Nazikian, G. J. Kramer, C. Z. Cheng, N. N. Gorelenkov, H. L. Berk, and S. E. Sharapov

2003-09-19T23:59:59.000Z

143

Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor  

SciTech Connect (OSTI)

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

Blaine Grover

2012-10-01T23:59:59.000Z

144

LANL demolishes first containment dome at disposal area  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

LANL Demolishes First Containment Dome LANL Demolishes First Containment Dome LANL demolishes first containment dome at disposal area It once housed thousands of drums of radioactive waste that have been shipped to the Waste Isolation Pilot Plant for disposal. September 30, 2009 Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials. Los Alamos National Laboratory sits on top of a once-remote mesa in northern New Mexico with the Jemez mountains as a backdrop to research and innovation covering multi-disciplines from bioscience, sustainable energy sources, to plasma physics and new materials.

145

A mechanical model of early salt dome growth  

E-Print Network [OSTI]

A MECHANICAL MODEL OF EARLY SALT DOME GROWTH A Thesis by FRANK ALBERT IRWIN Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE December 1988... Major Subject: Geology A MECHANICAL MODEL OF EARLY SALT DOME GROWTH A Thesis by FRANK ALBERT IRWIN Approved as to style and content by: aymond C. Fletcher (Chair of Committee) John H. Spang (Member) Wi tamR. B ant (Mem ) John H. Sp g (Head...

Irwin, Frank Albert

2012-06-07T23:59:59.000Z

146

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters  

SciTech Connect (OSTI)

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Konzek, G.J.

1983-07-01T23:59:59.000Z

147

Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors  

SciTech Connect (OSTI)

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes the fabrication and testing of full-length prototypcial converters, both unfueled and fueled, and presents parametric results of electrically heated tests.

Schock, Alfred

1995-08-01T23:59:59.000Z

148

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2  

SciTech Connect (OSTI)

A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

1981-09-01T23:59:59.000Z

149

The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures  

E-Print Network [OSTI]

shaft. With the impeller in place and rotating, gas was drawn into the top port and ejected at the impeller mount. The reactor pressure was monitored via the transducer port. The transducer was a Viatran Pressure Transducer, model 103. The liquid...THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial...

Hess, Richard Kenneth

2012-06-07T23:59:59.000Z

150

Direct Test of the Time-Independence of Fundamental Nuclear Constants Using the Oklo Natural Reactor  

E-Print Network [OSTI]

[NOTE: This 1983 preprint is being uploaded to arXiv.org after the death of its author, who supported online distribution of his work. Contact info of the submitter is at http://ilya.cc .] The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

Alexander I. Shlyakhter

2003-07-03T23:59:59.000Z

151

Effects of 50/degree/C surveillance and test reactor irradiations on ferritic pressure vessel steel embrittlement  

SciTech Connect (OSTI)

The results of surveillance tests on the High-Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory revealed that a greater than expected embrittlement had taken place after about 17.5 effective full-power years of operation and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed. A research program was undertaken that included irradiating specimens in the Oak Ridge Research Reactor. Specimens of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged-arc weld fabricated at ORNL to reproduce the vessel seam weld. The results of the surveillance program and the materials research program performed in support of the evaluation of the HFIR pressure vessel are presented and show the welds to be more radiation resistant than the A212B. Results of irradiated tensile and annealing experiments are described as well as a discussion of mechanisms which may be responsible for enhanced hardening at low damage rates. 20 refs., 22 figs., 5 tabs.

Nanstad, R.K.; Iskander, S.K.; Rowcliffe, A.F.; Corwin, W.R.; Odette, G.R.

1988-01-01T23:59:59.000Z

152

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor  

SciTech Connect (OSTI)

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

C. B. Davis

2006-07-01T23:59:59.000Z

153

Calculated and measured gas formation in beryllium samples irradiated in the high flux materials testing reactor BR2  

SciTech Connect (OSTI)

Beryllium samples have been irradiated in BR2, the materials testing reactor of the Nuclear Research Centre SCK/CEN at Mol, Belgium, up to fission fluence values of 5.2 10{sup 22} n/cm{sup 2} at low temperature. The gas formation (helium, tritium), as measured by SCK/CEN, as well as the induced swelling of the beryllium samples and the enhancement of the swelling due to annealing have been presented at the 17th SOFT Conference (Rome, 14--18 Sept., 1992). Since this conference, helium measurements on the same samples have been carried out at RI and calculations of the gas production have been performed, taking into account the various formation schemes. The experimental results from SCK/CEN and from RI are compared with the calculated gas formations.

De Raedt, C.M.; Sannen, L.F.; Vanmechelen, P.J. [SCK/CEN, Mol (Belgium); Oliver, B.M. [Rockwell International Corp., Canoga Park, CA (United States). Rocketdyne Div.

1994-12-31T23:59:59.000Z

154

Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene  

SciTech Connect (OSTI)

The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10?MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10?MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

Zhao, M. L., E-mail: zml812@mail.ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Chen, Y. P.; Li, G. Q.; Luo, Z. P. [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China)] [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Guo, H. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); General Atomics, P.O. Box 85608, San Diego, California 92186 (United States); Ye, M. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Tendler, M. [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)] [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)

2014-05-15T23:59:59.000Z

155

Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables  

SciTech Connect (OSTI)

U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

Lisa Harvego; Brion Bennett

2011-11-01T23:59:59.000Z

156

Nondestructive Testing of NITE-SiC Ceramics for Fusion Reactor Application  

Science Journals Connector (OSTI)

Ultrasonic test (UT) method is a suitable one for examining the inherent structural defects such as pores and cracks of SiC/SiC composites. In this study, the examination sensitivity limit of UT method based o...

Yun-Seok Shin; Yi-Hyun Park; Tatsuya Hinoki

2010-01-01T23:59:59.000Z

157

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect (OSTI)

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, S.B. (INEEL); Metcalfe, M.P. (BNFL, United Kingdom)

2002-03-14T23:59:59.000Z

158

Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors  

SciTech Connect (OSTI)

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment.

Grover, Stanley Blaine; Metcalfe, M. P.

2002-04-01T23:59:59.000Z

159

Geological evaluation of Gulf Coast salt domes: overall assessment of the Gulf Interior Region  

SciTech Connect (OSTI)

The three major phases in site characterization and selection are regional studies, area studies, and location studies. This report characterizes regional geologic aspects of the Gulf Coast salt dome basins. It includes general information from published sources on the regional geology; the tectonic, domal, and hydrologic stability; and a brief description the salt domes to be investigated. After a screening exercise, eight domes were chosen for further characterization: Keechi, Oakwood, and Palestine Domes in Texas; Vacherie and Rayburn's domes in North Louisiana; and Cypress Creek and Richton domes in Mississippi. A general description of each, maps of the location, property ownership, and surface geology, and a geologic cross section were presented for each dome.

none,

1981-10-01T23:59:59.000Z

160

A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE  

SciTech Connect (OSTI)

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

Jorge Navarro

2013-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Tatum Dome field study and analysis of monitoring data  

SciTech Connect (OSTI)

This report is a summary and interpretation of data collected at the Tatum Dome Site during April and May 1984. In addition to raw data, description of field activities and laboratory analytical results are presented. The report also includes all analytical results of shallow well samples from the site, and chemical, mineralogical and mechanical analyses of soil samples collected during April 1985. Analysis and interpretation of laboratory analyses are included.

Fordham, J.W.; Fenske, P.R.

1985-12-01T23:59:59.000Z

162

Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2  

SciTech Connect (OSTI)

A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

1981-09-01T23:59:59.000Z

163

Testing the Reactor and Gallium Anomalies with Intense (Anti)Neutrino Emitters  

E-Print Network [OSTI]

Several observed anomalies in neutrino oscillation data could be explained by a hypothetical fourth neutrino separated from the three standard neutrinos by a squared mass difference of a few 0.1 eV$^2$ or more. This hypothesis can be tested with MCi neutrino electron capture sources ($^{51}$Cr) or kCi antineutrino $\\beta$-source ($^{144}$Ce) deployed inside or next to a large low background neutrino detector. In particular, the compact size of this source coupled with the localization of the interaction vertex lead to an oscillating pattern in event spatial (and possibly energy) distributions that would unambiguously determine neutrino mass differences and mixing angles.

Th. Lasserre

2012-09-23T23:59:59.000Z

164

Status of the Norwegian thorium light water reactor (LWR) fuel development and irradiation test program  

SciTech Connect (OSTI)

Thorium based fuels offer several benefits compared to uranium based fuels and should thus be an attractive alternative to conventional fuel types. In order for thorium based fuel to be licensed for use in current LWRs, material properties must be well known for fresh as well as irradiated fuel, and accurate prediction of fuel behavior must be possible to make for both normal operation and transient scenarios. Important parameters are known for fresh material but the behaviour of the fuel under irradiation is unknown particularly for low Th content. The irradiation campaign aims to widen the experience base to irradiated (Th,Pu)O{sub 2} fuel and (Th,U)O{sub 2} with low Th content and to confirm existing data for fresh fuel. The assumptions with respect to improved in-core fuel performance are confirmed by our preliminary irradiation test results, and our fuel manufacture trials so far indicate that both (Th,U)O{sub 2} and (Th,Pu)O{sub 2} fuels can be fabricated with existing technologies, which are possible to upscale to commercial volumes.

Drera, S.S.; Bjork, K.I.; Kelly, J.F.; Asphjell, O. [Thor Energy AS: Sommerrogaten 13-15, Oslo, NO255 (Norway)

2013-07-01T23:59:59.000Z

165

Conceptual model for regional radionuclide transport from a salt dome repository: a technical memorandum  

SciTech Connect (OSTI)

Disposal of high-level radioactive wastes is a major environmental problem influencing further development of nuclear energy in this country. Salt domes in the Gulf Coast Basin are being investigated as repository sites. A major concern is geologic and hydrologic stability of candidate domes and potential transport of radionuclides by groundwater to the biosphere prior to their degradation to harmless levels of activity. This report conceptualizes a regional geohydrologic model for transport of radionuclides from a salt dome repository. The model considers transport pathways and the physical and chemical changes that would occur through time prior to the radionuclides reaching the biosphere. Necessary, but unknown inputs to the regional model involve entry and movement of fluids through the repository dome and across the dome-country rock interface and the effect on the dome and surrounding strata of heat generated by the radioactive wastes.

Kier, R.S.; Showalter, P.A.; Dettinger, M.D.

1980-05-30T23:59:59.000Z

166

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2012-02-01T23:59:59.000Z

167

Predicting optical and thermal characteristics of transparent single-glazed domed skylights  

SciTech Connect (OSTI)

Optical and thermal characteristics of domed skylights are important to solve the trade-off between daylighting and thermal design. However, there is a lack of daylighting and thermal design tools for domed skylights. Optical and thermal characteristics of transparent single-glazed hemispherical domed skylights under sun and sky light are evaluated based on an optical model for domed skylights. The optical model is based on tracing the beam and diffuse radiation transmission through the dome surface. A simple method is proposed to replace single-glazed hemispherical domed skylights by optically and thermally equivalent single-glazed planar skylights to accommodate limitations of energy computer programs. Under sunlight, single-glazed hemispherical domed skylights yield slightly lower equivalent solar transmittance and solar heat gain coefficient (SHGC) at near normal zenith angles than those of single-glazed planar skylights. However, single-glazed hemispherical domed skylights yield substantially higher equivalent solar transmittance and SHGC at high zenith angles and around the horizon. Under isotropic skylight, single-glazed hemispherical domed skylights yield slightly lower equivalent solar transmittance and SHGC than those of single-glazed planar skylights. Daily solar heat gains of single-glazed hemispherical domed skylights are higher than those of single-glazed horizontal planar skylights in both winter and summer. In summer, the solar heat gain of single-glazed hemispherical domed skylights can reach 3% to 9% higher than those of horizontal single-glazed planar skylights for latitudes varying between 0 and 55{degree} (north/south). In winter, however, the solar heat gains of single-glazed hemispherical domed skylights increase significantly with the increase of the site latitude and can reach 232% higher than those of horizontal single-glazed planar skylights, particularly for high latitude countries.

Laouadi, A.; Atif, M.R.

1999-07-01T23:59:59.000Z

168

Fractures of the Dammam Dome Carbonate Outcrops: Their Characterization, Development, and Implications for Subsurface Reservoirs.  

E-Print Network [OSTI]

??The exposed Tertiary carbonates of the Dammam Dome present an opportunity to study fractures in outcrops within the oil-producing region of Eastern Saudi Arabia. The… (more)

Al-Fahmi, Mohammed M

2012-01-01T23:59:59.000Z

169

Let’s Try That Again: Selling the Teapot Dome Oil Field  

Broader source: Energy.gov [DOE]

The first time the Teapot Dome Oilfield was sold, it threw the Harding administration into scandal. Now -- 93 years later -- we're selling the field legally.

170

Draft environmental assessment: Richton Dome site, Mississippi. Nuclear Waste Policy Act (Section 112). [Contains Glossary  

SciTech Connect (OSTI)

In February 1983, the US Department of Energy identified the Richton dome site as one of the nine potentially acceptable sites for a mined geo

Not Available

1984-12-01T23:59:59.000Z

171

Nuclear Reactor (atomic reactor)  

Science Journals Connector (OSTI)

A nuclear reactor splits Uranium or Plutonium nuclei, and the...235 is fissionable but more than 99% of the naturally occurring Uranium is U238 that makes enrichment mandatory. In some reactors U238 and Thorium23...

2008-01-01T23:59:59.000Z

172

Dome-Tech Merck Teaming Profile Pres 3-17-05  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

, 2005 , 2005 Energy Conservation Through Retro- Commissioning of Building 800 Presented by: TM , CEM, NEBB Vincent Gates, Energy Manager, Merck Keith A. Rinaldi, LEED 2 10-Mar-05 Dome-Tech Field Services We specialize in "hands on" testing, evaluation and Retro-Cx of existing HVAC and utility systems. We identify and solve problems, reduce energy expenses and improve indoor air quality. We save companies significant money without requiring large capital outlays. 3 10-Mar-05 Building 800 - Background * Multi-Science Building * Constructed and occupied in 2000 * 325,000 square feet * Consists of R&D laboratories, support spaces and administrative offices * Sixteen air handling units with VFD's * Sixteen constant volume exhaust fans * Three 200 HP chilled water pumps

173

Radar investigation of the Cote Blanche salt dome  

E-Print Network [OSTI]

PADAR IiVVESTICATIOH CF THE COZE ELANCHR SALT DO&: A Thests EOHEET DOSAED SZEVAET Subm-'tt. ii' to the Crsduete Coilege of Ter: s AVi', i;nlu xsity in partfal fulfi' line?t of th ': quiremen fc z' tht degree o %P t S "t Clt. 'iCE iugust. l...HIC major Sub jest: Ceoohysfes RADAR INVESTIGATION OP THE COTE BLANCHE SALT DOME A Thesis by ROBERT DONALD STEWART Approved as to style and nor. tent by: (Chairman of Comml ee) ( Ir~c" (Head of De rtment ? Member) (Member) August 1974 ABSTRACT...

Stewart, Robert Donald

2012-06-07T23:59:59.000Z

174

03/01/2006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http://english.people.com.cn/200603/01/print20060301_247035.html  

E-Print Network [OSTI]

03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion and the former Soviet Union launched a 10 billion- euro ambitious plan, the International Thermonuclear

175

Fast foldable tent domes Aswin P.L. Jgers a,b  

E-Print Network [OSTI]

-gust accelerations around large obstacles. This applies also to future large solar telescopes. At present two over the whole dome size. Simultaneously, a variety of wind-speed and -direction sensors measure the wind field around the dome. In addition, fast sensitive air- pressure sensors placed on the supporting

Rutten, Rob

176

Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From  

Open Energy Info (EERE)

Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Details Activities (5) Areas (1) Regions (0) Abstract: Temperatures of 100°C are measured at 3 km depth in a well located on the resurgent dome in the center of Long Valley Caldera, California, despite an assumed >800°C magma chamber at 6-8 km depth. Local downflow of cold meteoric water as a process for cooling the resurgent dome is ruled out by a Peclet-number analysis of temperature logs. These analyses reveal zones with fluid circulation at the upper and lower

177

Sensitivity of storage field performance to geologic and cavern design parameters in salt domes.  

SciTech Connect (OSTI)

A sensitivity study was performed utilizing a three dimensional finite element model to assess allowable cavern field sizes for strategic petroleum reserve salt domes. A potential exists for tensile fracturing and dilatancy damage to salt that can compromise the integrity of a cavern field in situations where high extraction ratios exist. The effects of salt creep rate, depth of salt dome top, dome size, caprock thickness, elastic moduli of caprock and surrounding rock, lateral stress ratio of surrounding rock, cavern size, depth of cavern, and number of caverns are examined numerically. As a result, a correlation table between the parameters and the impact on the performance of storage field was established. In general, slower salt creep rates, deeper depth of salt dome top, larger elastic moduli of caprock and surrounding rock, and a smaller radius of cavern are better for structural performance of the salt dome.

Ehgartner, Brian L. (Sandia National Laboratories, Albuquerque, NM); Park, Byoung Yoon

2009-03-01T23:59:59.000Z

178

Sensitivity of storage field performance to geologic and cavern design parameters in salt domes.  

SciTech Connect (OSTI)

A sensitivity study was performed utilizing a three dimensional finite element model to assess allowable cavern field sizes in strategic petroleum reserve salt domes. A potential exists for tensile fracturing and dilatancy damage to salt that can compromise the integrity of a cavern field in situations where high extraction ratios exist. The effects of salt creep rate, depth of salt dome top, dome size, caprock thickness, elastic moduli of caprock and surrounding rock, lateral stress ratio of surrounding rock, cavern size, depth of cavern, and number of caverns are examined numerically. As a result, a correlation table between the parameters and the impact on the performance of a storage field was established. In general, slower salt creep rates, deeper depth of salt dome top, larger elastic moduli of caprock and surrounding rock, and a smaller radius of cavern are better for structural performance of the salt dome.

Ehgartner, Brian L.; Park, Byoung Yoon; Herrick, Courtney Grant

2010-06-01T23:59:59.000Z

179

Strategic Petroleum Reserve (SPR) geological site characterization report, Big Hill Salt Dome  

SciTech Connect (OSTI)

Geological and geophysical analyses of the Big Hill Salt Dome were performed to determine the suitability of this site for use in the Strategic Petroleum Reserve (SPR). Development of 140 million barrels (MMB) of storage capacity in the Big Hill Salt Dome is planned as part of the SPR expansion to achieve 750 MMB of storage capacity. Objectives of the study were to: (1) Acquire, evaluate, and interpret existing data pertinent to geological characterization of the Big Hill Dome; (2) Characterize the surface and near-surface geology and hydrology; (3) Characterize the geology and hydrology of the overlying cap rock; (4) Define the geometry and geology of the dome; (5) Determine the feasibility of locating and constructing 14 10-MMB storage caverns in the south portion of the dome; and (6) Assess the effects of natural hazards on the SPR site. Recommendations are included. (DMC)

Hart, R.J.; Ortiz, T.S.; Magorian, T.R.

1981-09-01T23:59:59.000Z

180

Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount  

Open Energy Info (EERE)

Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount St Helens, Washington Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: Vlf Electromagnetic Investigations Of The Crater And Central Dome Of Mount St Helens, Washington Details Activities (1) Areas (1) Regions (0) Abstract: A very low frequency (VLF) electromagnetic induction survey in the crater of Mount St. Helens has identified several electrically conductive structures that appear to be associated with thermal anomalies and ground water within the crater. The most interesting of these conductive structures lies beneath the central dome. It is probably a partial melt of dacite similar to that comprising the June 1981 lobe of the central dome. Author(s): James N. Towle

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect (OSTI)

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

182

Design and testing of a boron carbide capsule for spectral-tailoring in mixed-spectrum reactors  

SciTech Connect (OSTI)

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State Univ.. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY'SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in {sup 10}B shows that it is possible to produce a neutron spectrum similar to that of {sup 235}U fission. (authors)

Greenwood, L.R.; Wittman, R. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Pierson, B.P. [Univ. of Michigan, Ann Arbor, MI 48109 (United States); Metz, L.A.; Payne, R.; Finn, E.C.; Friese, J.I. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States)

2011-07-01T23:59:59.000Z

183

Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425  

SciTech Connect (OSTI)

The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)

Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Stanke, Dieter [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)] [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)

2013-07-01T23:59:59.000Z

184

Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)  

SciTech Connect (OSTI)

The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

2012-07-01T23:59:59.000Z

185

Reactor Safety Planning for Prometheus Project, for Naval Reactors Information  

SciTech Connect (OSTI)

The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

P. Delmolino

2005-05-06T23:59:59.000Z

186

Time of growth of Opelika Dome, Henderson and Van Zandt counties, Texas  

E-Print Network [OSTI]

ABSTRACT The Gpelilca fielc! is located in lienderson nuM Van Zandt countiss? Texas. This location is ir. the western part c f the petroliferous F~t Texas Basin. The fielo has a dome-shapeu stxucture wl&ich is believed to hove been caused bi' a &esp...-seated salt ccome. At least 27 salt domes have been identifiec! in the Fast Texas Basin ared the;-. tructures of the known salt doves ars similar to the structure of the Opelika dome. The times and amounts cf, &cx&wth of ths & ore as well as the tires...

Peterson, Thomas Lowe

2012-06-07T23:59:59.000Z

187

Experimental investigations on decay heat removal in advanced nuclear reactors using single heater rod test facility: Air alone in the annular gap  

SciTech Connect (OSTI)

During a loss of coolant accident in nuclear reactors, radiation heat transfer accounts for a significant amount of the total heat transfer in the fuel bundle. In case of heavy water moderator nuclear reactors, the decay heat of a fuel bundle enclosed in the pressure tube and outer concentric calandria tube can be transferred to the moderator. Radiation heat transfer plays a significant role in removal of decay heat from the fuel rods to the moderator, which is available outside the calandria tube. A single heater rod test facility is designed and fabricated as a part of preliminary investigations. The objective is to anticipate the capability of moderator to remove decay heat, from the reactor core, generated after shut down. The present paper focuses mainly on the role of moderator in removal of decay heat, for situation with air alone in the annular gap of pressure tube and calandria tube. It is seen that the naturally aspirated air is capable of removing the heat generated in the system compared to the standstill air or stagnant water situations. It is also seen that the flowing moderator is capable of removing a greater fraction of heat generated by the heater rod compared to a stagnant pool of boiling moderator. (author)

Bopche, Santosh B.; Sridharan, Arunkumar [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400 076 (India)

2010-11-15T23:59:59.000Z

188

Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A  

SciTech Connect (OSTI)

Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

2009-11-30T23:59:59.000Z

189

Structural stability of andesite volcanoes and lava domes  

Science Journals Connector (OSTI)

...limit equilibrium analysis (LEA) methods...high degree of reliability. The principles...quantitative stability analyses developed. It...highly uncertain reliability. By mid-May...Finite element analysis of deformation...Structural Mechanics Reactor Technology, Chicago...

2000-01-01T23:59:59.000Z

190

The Thermal Regime In The Resurgent Dome Of Long Valley Caldera,  

Open Energy Info (EERE)

Thermal Regime In The Resurgent Dome Of Long Valley Caldera, Thermal Regime In The Resurgent Dome Of Long Valley Caldera, California- Inferences From Precision Temperature Logs In Deep Wells Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Journal Article: The Thermal Regime In The Resurgent Dome Of Long Valley Caldera, California- Inferences From Precision Temperature Logs In Deep Wells Details Activities (1) Areas (1) Regions (0) Abstract: Long Valley Caldera in eastern California formed 0.76 Ma ago in a cataclysmic eruption that resulted in the deposition of 600 km3 of Bishop Tuff. The total current heat flow from the caldera floor is estimated to be ~ 290 MW, and a geothermal power plant in Casa Diablo on the flanks of the resurgent dome (RD) generates ~40 MWe. The RD in the center of the caldera was uplifted by ~ 80 cm between 1980 and 1999 and was explained by most

191

EIS-0010: Strategic Petroleum Reserves, Sulphur Mines Salt Dome, Calcasieu Parish, Louisiana  

Broader source: Energy.gov [DOE]

The Strategic Petroleum Reserves prepared this EIS to assess the environmental impacts of the proposed storage of 24 million barrels of crude oil at the Sulphur Mines salt dome located in Calcasieu Parish, Louisiana, including construction and operation impacts.

192

Structural constraints on the exhumation of the Tso Morari Dome, NW Himalaya  

E-Print Network [OSTI]

The Tso Morari culmination in the Ladakh region of northwest India is a large (>3,000 km²) structural dome cored by coesite-bearing rocks of Indian continental crustal affinity. As one of only two localities in the Himalaya ...

Clark, Ryan J

2005-01-01T23:59:59.000Z

193

Fluid Flow In The Resurgent Dome Of Long Valley Caldera- Implications...  

Open Energy Info (EERE)

The Resurgent Dome Of Long Valley Caldera- Implications From Thermal Data And Deep Electrical Sounding Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal...

194

Variability of bottom water domes and geostrophic currents in the eastern Gulf of Maine  

E-Print Network [OSTI]

VARIABILITY OF BOTTOM WATER DOMES AND GEOSTROPHIC CURRENTS IN THE EASTERN GULF OF MAINE A Thesis by ERIK SAUL GQTTLIEB Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree... of MASTER OF SCIENCE December 1987 Major Subject: Oceanography VARIABILITY OF BOTTOM WATER DOMES AND GEOSTROPHIC CURRENTS IN THE EASTERN GULF OF ~ A Thesis by ERIK SAUL GOTTLIEB Approved as to style and content by: Davi A. rooks (Chairman...

Gottlieb, Erik S

2012-06-07T23:59:59.000Z

195

Three-dimensional representations of salt-dome margins at four active strategic petroleum reserve sites.  

SciTech Connect (OSTI)

Existing paper-based site characterization models of salt domes at the four active U.S. Strategic Petroleum Reserve sites have been converted to digital format and visualized using modern computer software. The four sites are the Bayou Choctaw dome in Iberville Parish, Louisiana; the Big Hill dome in Jefferson County, Texas; the Bryan Mound dome in Brazoria County, Texas; and the West Hackberry dome in Cameron Parish, Louisiana. A new modeling algorithm has been developed to overcome limitations of many standard geological modeling software packages in order to deal with structurally overhanging salt margins that are typical of many salt domes. This algorithm, and the implementing computer program, make use of the existing interpretive modeling conducted manually using professional geological judgement and presented in two dimensions in the original site characterization reports as structure contour maps on the top of salt. The algorithm makes use of concepts of finite-element meshes of general engineering usage. Although the specific implementation of the algorithm described in this report and the resulting output files are tailored to the modeling and visualization software used to construct the figures contained herein, the algorithm itself is generic and other implementations and output formats are possible. The graphical visualizations of the salt domes at the four Strategic Petroleum Reserve sites are believed to be major improvements over the previously available two-dimensional representations of the domes via conventional geologic drawings (cross sections and contour maps). Additionally, the numerical mesh files produced by this modeling activity are available for import into and display by other software routines. The mesh data are not explicitly tabulated in this report; however an electronic version in simple ASCII format is included on a PC-based compact disk.

Rautman, Christopher Arthur; Stein, Joshua S.

2003-01-01T23:59:59.000Z

196

Design and Testing of a 10B4C Capsule for Spectral-Tailoring in Mixed-Spectrum Reactors  

SciTech Connect (OSTI)

A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum. A cadmium outer cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

Greenwood, Lawrence R.; Wittman, Richard S.; Metz, Lori A.; Finn, Erin C.; Friese, Judah I.

2014-04-11T23:59:59.000Z

197

Features of Bayou Choctaw SPR caverns and internal structure of the salt dome.  

SciTech Connect (OSTI)

The intent of this study is to examine the internal structure of the Bayou Choctaw salt dome utilizing the information obtained from graphical representations of sonar survey data of the internal cavern surfaces. Many of the Bayou Choctaw caverns have been abandoned. Some existing caverns were purchased by the Strategic Petroleum Reserve (SPR) program and have rather convoluted histories and complex cavern geometries. In fact, these caverns are typically poorly documented and are not particularly constructive to this study. Only two Bayou Choctaw caverns, 101 and 102, which were constructed using well-controlled solutioning methods, are well documented. One of these was constructed by the SPR for their use while the other was constructed and traded for another existing cavern. Consequently, compared to the SPR caverns of the West Hackberry and Big Hill domes, it is more difficult to obtain a general impression of the stratigraphy of the dome. Indeed, caverns of Bayou Choctaw show features significantly different than those encountered in the other two SPR facilities. In the number of abandoned caverns, and some of those existing caverns purchased by the SPR, extremely irregular solutioning has occurred. The two SPR constructed caverns suggest that some sections of the caverns may have undergone very regular solutioning to form uniform cylindrical shapes. Although it is not usually productive to speculate, some suggestions that point to the behavior of the Bayou Choctaw dome are examined. Also the primary differences in the Bayou Choctaw dome and the other SPR domes are noted.

Munson, Darrell E.

2007-07-01T23:59:59.000Z

198

A comparison between semi-spheroid- and dome-shaped quantum dots coupled to wetting layer  

SciTech Connect (OSTI)

During the epitaxial growth method, self-assembled semi-spheroid-shaped quantum dots (QDs) are formed on the wetting layer (WL). However for sake of simplicity, researchers sometimes assume semi-spheroid-shaped QDs to be dome-shaped (hemisphere). In this work, a detailed and comprehensive study on the difference between electronic and transition properties of dome- and semi-spheroid-shaped quantum dots is presented. We will explain why the P-to-S intersubband transition behaves the way it does. The calculated results for intersubband P-to-S transition properties of quantum dots show two different trends for dome-shaped and semi-spheroid-shaped quantum dots. The results are interpreted using the probability of finding electron inside the dome/spheroid region, with emphasis on the effects of wetting layer. It is shown that dome-shaped and semi-spheroid-shaped quantum dots feature different electronic and transition properties, arising from the difference in lateral dimensions between dome- and semi-spheroid-shaped QDs. Moreover, an analogy is presented between the bound S-states in the quantum dots and a simple 3D quantum mechanical particle in a box, and effective sizes are calculated. The results of this work will benefit researchers to present more realistic models of coupled QD/WL systems and explain their properties more precisely.

Shahzadeh, Mohammadreza; Sabaeian, Mohammad, E-mail: Sabaeian@scu.ac.ir [Physics Department, Faculty of Science, Shahid Chamran University of Ahvaz, Ahvaz, 61357-43135 (Iran, Islamic Republic of)

2014-06-15T23:59:59.000Z

199

Plasma–wall interaction in laser inertial fusion reactors: novel proposals for radiation tests of first wall materials  

Science Journals Connector (OSTI)

Dry-wall laser inertial fusion (LIF) chambers will have to withstand strong bursts of fast charged particles which will deposit tens of kJ m?2 and implant more than 1018 particles m?2 in a few microseconds at a repetition rate of some Hz. Large chamber dimensions and resistant plasma-facing materials must be combined to guarantee the chamber performance as long as possible under the expected threats: heating, fatigue, cracking, formation of defects, retention of light species, swelling and erosion. Current and novel radiation resistant materials for the first wall need to be validated under realistic conditions. However, at present there is a lack of facilities which can reproduce such ion environments.This contribution proposes the use of ultra-intense lasers and high-intense pulsed ion beams (HIPIB) to recreate the plasma conditions in LIF reactors. By target normal sheath acceleration, ultra-intense lasers can generate very short and energetic ion pulses with a spectral distribution similar to that of the inertial fusion ion bursts, suitable to validate fusion materials and to investigate the barely known propagation of those bursts through background plasmas/gases present in the reactor chamber. HIPIB technologies, initially developed for inertial fusion driver systems, provide huge intensity pulses which meet the irradiation conditions expected in the first wall of LIF chambers and thus can be used for the validation of materials too.

J Alvarez Ruiz; A Rivera; K Mima; D Garoz; R Gonzalez-Arrabal; N Gordillo; J Fuchs; K Tanaka; I Fernández; F Briones; J Perlado

2012-01-01T23:59:59.000Z

200

RMOTC to Test Oil Viscosity Reduction Technology  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

to Test Oil Viscosity Reduction Technology to Test Oil Viscosity Reduction Technology The Rocky Mountain Oilfield Testing Center (RMOTC) announces that the "Teapot Dome" oil field in Wyoming is hosting a series of tests funded by STWA, Inc. ("STWA") to determine the performance of its Applied Oil Technology (AOT(tm)) in reducing crude oil's viscosity to lower transportation costs for pipeline operators. The testing is managed by RMOTC, and conducted at Naval Petroleum Reserve No. 3, also known as the Teapot Dome oil field. RMOTC is providing the infrastructure and technical expertise to support companies such as STWA in their efforts to validate new technologies and bring those products and

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201

A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions  

SciTech Connect (OSTI)

A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup ?4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

Morfin, Franck; Piccolo, Laurent [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)] [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

2013-09-15T23:59:59.000Z

202

Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop  

SciTech Connect (OSTI)

The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

McCulloch, R.W.; MacPherson, R.E.

1983-03-01T23:59:59.000Z

203

Reactor Safety Research Programs  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1981-07-01T23:59:59.000Z

204

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

behavior in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and...

205

Degradation of Dome Cutting Minerals in Hanford Waste - 13100  

SciTech Connect (OSTI)

At the Hanford Tank Farms, recent changes in retrieval technology require cutting new risers in several single-shell tanks. The Hanford Tank Farm Operator is using water jet technology with abrasive silicate minerals such as garnet or olivine to cut through the concrete and rebar dome. The abrasiveness of these minerals, which become part of the high-level waste stream, may enhance the erosion of waste processing equipment. However, garnet and olivine are not thermodynamically stable in Hanford waste, slowly degrading over time. How likely these materials are to dissolve completely in the waste before the waste is processed in the Waste Treatment and Immobilization Plant can be evaluated using theoretical analysis for olivine and collected direct experimental evidence for garnet. Based on an extensive literature study, a large number of primary silicates decompose into sodalite and cancrinite when exposed to Hanford waste. Given sufficient time, the sodalite also degrades into cancrinite. Even though cancrinite has not been directly added to any Hanford tanks during process times, it is the most common silicate observed in current Hanford waste. By analogy, olivine and garnet are expected to ultimately also decompose into cancrinite. Garnet used in a concrete cutting demonstration was immersed in a simulated supernate representing the estimated composition of the liquid retrieving waste from Hanford tank 241-C-107 at both ambient and elevated temperatures. This simulant was amended with extra NaOH to determine if adding caustic would help enhance the degradation rate of garnet. The results showed that the garnet degradation rate was highest at the highest NaOH concentration and temperature. At the end of 12 weeks, however, the garnet grains were mostly intact, even when immersed in 2 molar NaOH at 80 deg. C. Cancrinite was identified as the degradation product on the surface of the garnet grains. In the case of olivine, the rate of degradation in the high-pH regimes of a waste tank is expected to depend on two main parameters: carbonate is expected to slow olivine degradation rates, whereas hydroxide is expected to enhance olivine dissolution rates. Which of these two competing dissolution drivers will have a larger impact on the dissolution rate in the specific environment of a waste tank is currently not identifiable. In general, cancrinite is much smaller and less hard than either olivine or garnet, so would be expected to be less erosive to processing equipment. Complete degradation of either garnet or olivine prior to being processed at the Waste Treatment and Immobilization Plant cannot be confirmed, however. (authors)

Reynolds, Jacob G.; Cooke, Gary A.; Huber, Heinz J. [Washington River Protection Solutions, LLC, P.O. Box 850, Richland, WA 99352 (United States)] [Washington River Protection Solutions, LLC, P.O. Box 850, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

206

Degradation of dome cutting minerals in Hanford waste  

SciTech Connect (OSTI)

At the Hanford Tank Farms, recent changes in retrieval technology require cutting new risers in several single-shell tanks. The Hanford Tank Farm Operator is using water jet technology with abrasive silicate minerals such as garnet or olivine to cut through the concrete and rebar dome. The abrasiveness of these minerals, which become part of the high-level waste stream, may enhance the erosion of waste processing equipment. However, garnet and olivine are not thermodynamically stable in Hanford waste, slowly degrading over time. How likely these materials are to dissolve completely in the waste before the waste is processed in the Waste Treatment and Immobilization Plant can be evaluated using theoretical analysis for olivine and collected direct experimental evidence for garnet. Based on an extensive literature study, a large number of primary silicates decompose into sodalite and cancrinite when exposed to Hanford waste. Given sufficient time, the sodalite also degrades into cancrinite. Even though cancrinite has not been directly added to any Hanford tanks during process times, it is the most common silicate observed in current Hanford waste. By analogy, olivine and garnet are expected to ultimately also decompose into cancrinite. Garnet used in a concrete cutting demonstration was immersed in a simulated supernate representing the estimated composition of the liquid retrieving waste from Hanford tank 241-C-107 at both ambient and elevated temperatures. This simulant was amended with extra NaOH to determine if adding caustic would help enhance the degradation rate of garnet. The results showed that the garnet degradation rate was highest at the highest NaOH concentration and temperature. At the end of 12 weeks, however, the garnet grains were mostly intact, even when immersed in 2 molar NaOH at 80 deg C. Cancrinite was identified as the degradation product on the surface of the garnet grains. In the case of olivine, the rate of degradation in the high-pH regimes of a waste tank is expected to depend on two main parameters: carbonate is expected to slow olivine degradation rates, whereas hydroxide is expected to enhance olivine dissolution rates. Which of these two competing dissolution drivers will have a larger impact on the dissolution rate in the specific environment of a waste tank is currently not identifiable. In general, cancrinite is much smaller and less hard than either olivine or garnet, so would be expected to be less erosive to processing equipment. Complete degradation of either garnet or olivine prior to being processed at the Waste Treatment and Immobilization Plant cannot be confirmed, however.

Reynolds, Jacob G.; Huber, Heinz J.; Cooke, Gary A.

2013-01-11T23:59:59.000Z

207

In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)  

SciTech Connect (OSTI)

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01T23:59:59.000Z

208

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered Three-dimensional Thermal and Airflow (3D-TAF) Model of a Dome-covered House in Canada Speaker(s): Yaolin Lin Date: October 6, 2009 - 12:00pm Location: 90-3122 A dome-covered house is an example of sustainable design that draws from biological forms in nature. A three-dimensional thermal and air flow (3D-TAF) model was developed to estimate the energy needs of a dome-covered house. This model has two components: a thermal model to calculate the temperature; and an air flow model to find the velocities, which are needed to estimate the surface convection. The two models are solved iteratively at every time step until they converge. I will present the numerical methods for solving the mathematical models, and compared the results with other simulated and experimental results from similar structures. I will

209

Dome-shaped microresonators and the Born-Oppenheimer Jens U. Nockela and David H. Fosterb  

E-Print Network [OSTI]

Dome-shaped microresonators and the Born-Oppenheimer method Jens U. N¨ockela and David H. Fosterb a explore the Born-Oppenheimer method as an alternative to the paraxial approximation. The conditions the major results of paraxial theory can also be derived from the Born-Oppenheimer method. We discuss how

Nöckelm, Jens

210

Reservoir simulation of co2 sequestration and enhanced oil recovery in Tensleep Formation, Teapot Dome field  

E-Print Network [OSTI]

Teapot Dome field is located 35 miles north of Casper, Wyoming in Natrona County. This field has been selected by the U.S. Department of Energy to implement a field-size CO2 storage project. With a projected storage of 2.6 million tons of carbon...

Gaviria Garcia, Ricardo

2006-04-12T23:59:59.000Z

211

Beryllium-10 in the Taylor Dome ice core: Applications to Antarctic glaciology and paleoclimatology  

SciTech Connect (OSTI)

An ice core was drilled at Taylor dome, East Antarctica, reaching to bedrock at 554 meters. Oxygen-isotope measurements reveal climatic fluctuations through the last interglacial period. To facilitate comparison of the Taylor Dome paleoclimate record with geologic data and results from other deep ice cores, several glaciological issues need to be addressed. In particular, accumulation data are necessary as input for numerical ice-flow-models, for determining the flux of chemical constituents from measured concentrations, and for calculation of the offset in age between ice and trapped air in the core. The analysis of cosmogenic beryllium-10 provides a geochemical method for constraining the accumulation-rate history at Taylor Dome. High-resolution measurements were made in shallow firn cores and snow pits to determine the relationship among beryllium-10 concentrations, wet and dry deposition mechanisms, and snow-accumulation rates. Comparison between theoretical and measured variations in deposition over the last 75 years constrains the relationship between beryllium-10 deposition and global average production rates. The results indicate that variations in geomagnetically-modulated production-rate do not strongly influence beryllium-10 deposition at Taylor Dome. Although solar modulation of production rate is important for time scales of years to centuries, snow-accumulation rate is the dominant control on ice-core beryllium-10 concentrations for longer periods. Results show that the Taylor Dome core can be used to provide new constraints on regional climate over the last 130,000 years, complementing the terrestrial and marine geological record from the Dry Valley, Transantarctic Mountains and western Ross Sea.

Steig, E.J.

1996-12-31T23:59:59.000Z

212

EA-0813; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL  

Broader source: Energy.gov (indexed) [DOE]

13; Environmental Assessment and (FONSI) The Tokamak Fusion 13; Environmental Assessment and (FONSI) The Tokamak Fusion Test Reactor Decontamination and Decommissioning Project and The Tokamak Physics Experiment at the PPPL Table of Contents EXECUTIVE SUMMARY ACRONYMS Glossary of Radiological Terms SCIENTIFIC NOTATION 1.0 PURPOSE AND NEED FOR THE PROPOSED ACTIONS 1.1 TFTR D&D Project 1.2 TPX Project 1.3 Scope of Document 1.4 Local Community Relations Program 1.5 References 2.0 DESCRIPTION OF THE PROPOSED ACTIONS AND ALTERNATIVES 2.1 TFTR D&D Project 2.2 TPX Project 2.3 Environmental Monitoring 2.4 References 3.0 DESCRIPTION OF THE AFFECTED ENVIRONMENT 3.1 PPPL Proposed Site 3.2 ORR Alternative Site 3.3 References 4.0 ENVIRONMENTAL CONSEQUENCES OF THE PROPOSED ACTIONS AND ALTERNATIVES 4.1 TFTR D&D Project 4.1.1 Impacts of Normal D&D Operations

213

A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94  

SciTech Connect (OSTI)

Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA).

B. R. Orr (USGS)

1999-11-01T23:59:59.000Z

214

2010 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2009 through October 31, 2010. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Discussion of the facility’s environmental impacts During the 2010 permit year, approximately 164 million gallons of wastewater were discharged to the Cold Waste Pond. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

mike lewis

2011-02-01T23:59:59.000Z

215

2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond  

SciTech Connect (OSTI)

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012–October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Noncompliance issues • Discussion of the facility’s environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Mike Lewis

2014-02-01T23:59:59.000Z

216

Public comment sought on final end state of Experimental Breeder Reactor-II  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Media Contacts: Danielle Miller, 208-526-5709 Media Contacts: Danielle Miller, 208-526-5709 Joseph Campbell, CWI, 208-360-0142 Public comment sought on final end state of Experimental Breeder Reactor-II The U.S. Department of Energy (DOE) is seeking public comment on a range of alternatives for disposition of the landmark Experimental Breeder Reactor-II (EBR-II) building and reactor vessel at the Idaho Site's Materials and Fuels Complex. An Engineering Evaluation/Cost Analysis (EE/CA) document with four proposed alternatives for the final end state of the reactor facility and support structures is currently under evaluation by DOE, the U.S. Environmental Protection Agency, and Idaho's Department of Environmental Quality. Experimental Breeder Reactor-II containment dome The EBR-II was an innovative sodium-cooled reactor with an output of 62

217

NUCLEAR REACTORS.  

E-Print Network [OSTI]

??Nuclear reactors are devices containing fissionable material in sufficient quantity and so arranged as to be capable of maintaining a controlled, self-sustaining NUCLEAR FISSION chain… (more)

Belachew, Dessalegn

2010-01-01T23:59:59.000Z

218

The origin of the structural depression above Gulf coast salt domes with particular reference to Clay Creek dome, Washington County, Texas  

E-Print Network [OSTI]

oT ~ttee THd ORIGTN OF Tiki STRUCTURAL S&iR "'SION ABOV' ULF COAST SALT KRISS UITH PARTI'ULAR ~~JCS TO ' LAY CRgdK DGIQ JA' HI:. ' TON COK;TY~ TWAS Alfred Norman licDowsll A Thesis Submitted to the Graduate School of the Agricultural... Application of kodel Theory to uepression problem 13 Salt mis 'cele Models 14 Hodel Fault Patterns . 16 Aegio. ial Setting of Slay Greek oalt Dome. . . ~ Structural Interpretation of Faulting on Glar Greek Tertiary Growth History of 'la; Greek Some...

McDowell, Alfred Norman

2012-06-07T23:59:59.000Z

219

Dome C for a GABLS4 experiment : a challenge ?  

E-Print Network [OSTI]

parameters (10 Hz) from 6 ultra-sonic anemometers : 3D Wind components and sonic temperature Low frequency and direction (Young) 1 minute solar radiation components Sub and surface temperatures Radiometer HAMSTRAD (P , time step=60s (SL), preliminary test with 60 and then 90 vertical levels ·Two types of experiment

Ribes, Aurélien

220

Nuclear Reactor Severe Accident Experiments  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Transport reactor development status  

SciTech Connect (OSTI)

This project is part of METC`s Power Systems Development Facility (PSDF) located at Wilsonville, Alabama. The primary objective of the Advanced Gasifier module is to produce vitiated gases for intermediate-term testing of Particulate Control Devices (PCDs). The Transport reactor potentially allows particle size distribution, solids loading, and particulate characteristics in the off-gas stream to be varied in a number of ways. Particulates in the hot gases from the Transport reactor will be removed in the PCDs. Two PCDs will be initially installed in the module; one a ceramic candle filter, the other a granular bed filter. After testing of the initial PCDs they will be removed and replaced with PCDs supplied by other vendors. A secondary objective is to verify the performance of a Transport reactor for use in advanced Integrated Gasification Combined Cycle (IGCC), Integrated Gasification Fuel Cell (IG-FC), and Pressurized Combustion Combined Cycle (PCCC) power generation units. This paper discusses the development of the Transport reactor design from bench-scale testing through pilot-scale testing to design of the Process Development Unit (PDU-scale) facility at Wilsonville.

Rush, R.E.; Fankhanel, M.O.; Campbell, W.M.

1994-10-01T23:59:59.000Z

222

NETL - Chemical Looping Reactor  

ScienceCinema (OSTI)

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2014-06-26T23:59:59.000Z

223

WCH Removes Massive Test Reactor  

Broader source: Energy.gov [DOE]

RICHLAND, WA -- Hanford's River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy's (DOE) Hanford Site by removing a 1,082...

224

Potassium-Modified Iron Ore as Oxygen Carrier for Coal Chemical Looping Combustion: Continuous Test in 1 kW Reactor  

Science Journals Connector (OSTI)

(20) Briefly, it consists of a fast fluidized bed as an air reactor, a cyclone, a spout-fluid bed as a fuel reactor, and an external loop-seal. ... Industrial & Engineering Chemistry Research (2013), 52 (28), 9573-9585 CODEN: IECRED; ISSN:0888-5885. ... Gu, H.; Shen, L.; Xiao, J.; Zhang, S.; Song, T.; Chen, D.Iron ore as oxygen carrier improved with potassium for chemical looping combustion of anthracite coal Combust. ...

Haiming Gu; Laihong Shen; Zhaoping Zhong; Xin Niu; Huijun Ge; Yufei Zhou; Shen Xiao

2014-07-30T23:59:59.000Z

225

High Flux Beam Reactor | Environmental Restoration Projects | BNL  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Complex Description Complex Description Current HFBR Complex The HFBR complex consists of multiple structures and systems that were necessary to operate and maintain the reactor. The most recognizable features of the complex are the domed reactor confinement building and the distinctive red-and-white stack. Portions of the complex building structures, systems, and components, some of which are underground, were contaminated with radionuclides and chemicals as a result of previous HFBR and Brookhaven Graphite Research Reactor (BGRR) operations. A number of decommissioning and preparation for long-term safe storage actions have been taken including the removal of contaminated structures, hazardous materials, and contaminated equipment and components. The structures and systems, both current and former, are

226

EIS-0029: Strategic Petroleum Reserve, Texoma Group Salt Domes, Cameron and Calcasieu Parishes, Louisiana and Jefferson County, TX  

Broader source: Energy.gov [DOE]

The Strategic Petroleum Reserves developed this EIS to analyze the environmental impacts which could occur during site preparation and operation of oil storage facilities at each of four proposed candidate sites in the Texoma Group of salt domes.

227

Seal integrity and feasibility of CO2 sequestration in the Teapot Dome EOR pilot: geomechanical site characterization  

Science Journals Connector (OSTI)

This paper reports a preliminary investigation of CO2 sequestration and seal integrity at Teapot Dome oil field, Wyoming, USA, with...2 leakage along reservoir-bounding faults. CO2 injection into reservoirs creat...

Laura Chiaramonte; Mark D. Zoback; Julio Friedmann; Vicki Stamp

2008-06-01T23:59:59.000Z

228

Salt tectonism and seismic stratigraphy of the Upper Jurassic in the Destin Dome Region, northeastern Gulf of Mexico  

E-Print Network [OSTI]

SALT TECI'ONISM AND SEISMIC STRATIGRAPHY OF THE UPPER JURASSIC IN THE DESTIN DOME REGION, NORTHEASTERN GULF OF MEXICO A Thesis by GRANT MACRAE Submitted to the Office of Graduate Studies of Texas A&M University in partial fulfillment... of the requirements for the degree of MASTER OF SCIENCE August 1990 Major Subject: Oceanography SALT TECI'ONISM AND SEISMIC STRATIGRAPHY OF THE UPPER JURASSIC IN THE DESTIN DOME REGION, NORTHEASTERN GULF OF MEXICO A Thesis by GRANT MACRAE Approved...

MacRae, Grant

2012-06-07T23:59:59.000Z

229

Use of dipmeter logs to refine structural mapping of Teapot Dome, Wyoming  

SciTech Connect (OSTI)

Teapot field, now Naval Petroleum Reserve 3, is an elongated, asymmetric structural dome with a north-northeast axial trend located on the southwest edge of the Powder River basin. Currently, more than 800 wells of various depths penetrate multiple reservoirs; over 30 dipmeter logs have been run during the past 34 years. Although structure contour maps of individual stratigraphic horizons have been drawn simply from conventional well data, more subtle features of deformation are interpreted from the use of dipmeter logs. Because dips are generally less than 15/sup 0/, careful computation of transverse and longitudinal dip directions was required for detailed structural analysis. All depth, dip, and azimuth data were entered into computer files, and a flow chart of steps for computer processing and structural interpretation was devised and followed. As designed by C.A. Bengston, SCAT (Statistical Curvature Analysis Technique) plots were drawn by computer. Interpretation of SCAT plots yielded quantitative descriptions of the asymmetry of Teapot dome, curvature of the anticlinal axial plane, vertical discontinuity of beds, location and orientation of normal faults, and curvature of beds in drag zones adjacent to faults. Structural definition was necessary to outline boundaries of reservoirs with tilted fluid contacts on the flanks of the dome and along fault planes. Location of such faults would be particularly important for the discovery of deeper pools, for instance, in the Tensleep Formation.

Beinkafner, K.J.

1986-08-01T23:59:59.000Z

230

naval reactors  

National Nuclear Security Administration (NNSA)

After operating for 34 years and training over 14,000 sailors, the Department of Energy S1C Prototype Reactor Site in Windsor, Connecticut, was returned to "green field"...

231

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

Schock, Alfred

1994-08-01T23:59:59.000Z

232

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect (OSTI)

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28T23:59:59.000Z

233

Assessment of Effectiveness of Geologic Isolation Systems: REFERENCE SITE INITIAL ASSESSMENT FOR A SALT DOME REPOSITORY  

SciTech Connect (OSTI)

As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario resulted in the delivery of radionuclidecontaminated brine to the surface, where a portion was diverted to culinary salt for direct ingestion by the existing population. Consequence analyses indicated calculated human doses that would be highly deleterious. Additional analyses indicated that doses well above background would occur from such a scenario t even if it occurred a million years into the future. The way to preclude such an intrusion is for continued control over the repository sitet either through direct institutional control or through the effective passive transfer of information. A secondary aspect of the specific human intrusion scenario involved a breach through the side of the salt dome t through which radionuclides migrated via the ground-water system to the accessible environment. This provided a demonstration of the geotransport methodology that AEGIS can use in actual site evaluations, as well as the WRIT program's capabilities with respect to defining the source term and retardation rates of the radionuclides in the repository. This reference site analysis was initially published as a Working Document in December 1979. That version was distributed for a formal peer review by individuals and organizations not involved in its development. The present report represents a revisiont based in part on the responses received from the external reviewers. Summaries of the comments from the reviewers and responses to these comments by the AEGIS staff are presented. The exercise of the AEGIS methodology was successful in demonstrating the methodologyt and thus t in providing a basis for substantive peer review, in terms of further development of the AEGIS site-applications capability and in terms of providing insight into the potential for consequential human intrusion into a salt dome repository.

Harwell,, M. A.; Brandstetter,, A.; Benson,, G. L.; Bradley,, D. J.; Serne,, R. J.; Soldat, J. K; Cole,, C. R.; Deutsch,, W. J.; Gupta,, S. K.; Harwell,, C. C.; Napier,, B. A.; Reisenauer,, A. E.; Prater,, L. S.; Simmons,, C. S.; Strenge,, D. L.; Washburn,, J. F.; Zellmer,, J. T.

1982-06-01T23:59:59.000Z

234

Assessment of Effectiveness of Geologic Isolation Systems: REFERENCE SITE INITIAL ASSESSMENT FOR A SALT DOME REPOSITORY  

SciTech Connect (OSTI)

As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario resulted in the delivery of radionuclidecontaminated brine to the surface, where a portion was diverted to culinary salt for direct ingestion by the existing population. Consequence analyses indicated calculated human doses that would be highly deleterious. Additional analyses indicated that doses well above background would occur from such a scenario t even if it occurred a million years into the future. The way to preclude such an intrusion is for continued control over the repository sitet either through direct institutional control or through the effective passive transfer of information. A secondary aspect of the specific human intrusion scenario involved a breach through the side of the salt dome t through which radionuclides migrated via the ground-water system to the accessible environment. This provided a demonstration of the geotransport methodology that AEGIS can use in actual site evaluations, as well as the WRIT program's capabilities with respect to defining the source term and retardation rates of the radionuclides in the repository. This reference site analysis was initially published as a Working Document in December 1979. That version was distributed for a formal peer review by individuals and organizations not involved in its development. The present report represents a revisiont based in part on the responses received from the external reviewers. Summaries of the comments from the reviewers and responses to these comments by the AEGIS staff are presented. The exercise of the AEGIS methodology was sUGcessful in demonstrating the methodologyt and thus t in providing a basis for substantive peer review, in terms of further development of the AEGIS site-applications capability and in terms of providing insight into the potential for consequential human intrusion into a salt dome repository.

Harwell,, M. A.; Brandstetter,, A.; Benson,, G. L.; Raymond,, J. R.; Brandley,, D. J.; Serne,, R. J.; Soldat,, J. K.; Cole,, C. R.; Deutsch,, W. J.; Gupta,, S. K.; Harwell,, C. C.; Napier,, B. A.; Reisenauer,, A. E.; Prater,, L. S.; Simmons,, C. S.; Strenge,, D. L.; Washburn,, J. F.; Zellmer,, J. T.

1982-06-01T23:59:59.000Z

235

Research reactors - an overview  

SciTech Connect (OSTI)

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

236

Light Water Reactor Sustainability  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

237

Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design  

SciTech Connect (OSTI)

Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation.

Chu, T.Y.; Bentz, J.H.; Simpson, R.B.

1995-06-01T23:59:59.000Z

238

Catalytic reactor  

DOE Patents [OSTI]

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

239

The Thermal Environment of the Fiber Glass Dome for the New Solar Telescope at Big Bear Solar Observatory  

E-Print Network [OSTI]

The New Solar Telescope (NST) is a 1.6-meter off-axis Gregory-type telescope with an equatorial mount and an open optical support structure. To mitigate the temperature fluctuations along the exposed optical path, the effects of local/dome-related seeing have to be minimized. To accomplish this, NST will be housed in a 5/8-sphere fiberglass dome that is outfitted with 14 active vents evenly spaced around its perimeter. The 14 vents house louvers that open and close independently of one another to regulate and direct the passage of air through the dome. In January 2006, 16 thermal probes were installed throughout the dome and the temperature distribution was measured. The measurements confirmed the existence of a strong thermal gradient on the order of 5 degree Celsius inside the dome. In December 2006, a second set of temperature measurements were made using different louver configurations. In this study, we present the results of these measurements along with their integration into the thermal control system (ThCS) and the overall telescope control system (TCS).

A. P. Verdoni; C. Denker; J. R. Varsik; S. Shumko; J. Nenow; R. Coulter

2007-08-04T23:59:59.000Z

240

Experience in testing of a solution mined storage cavern  

SciTech Connect (OSTI)

Recertification tests were made of the U.S. Department of Energy/Strategic Petroleum Reserve oil storage cavern No. 6 in the West Hackberry, LA, salt dome. The cavern has a volume of 8,600,000 bbl. Tests included hydrostatic tests of the brine filled cavern and nitrogen leak tests of the 3 wells entering the cavern. Test procedures are described and test results are discussed.

Goin, K.L.

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Development and validation of a real-time SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) system for the inspection of light water reactor components: Annual report, October 1985-September 1986  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory is working to design, fabricate, and evaluate a real-time flaw detection and characterization system based on the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT). The system is designed to perform inservice inspection of light-water reactor components. Included objectives of this program for the Nuclear Regulatory Commission are to develop procedures for system calibration and field operation, to validate the system through laboratory and field inspections, and to generate an engineering data base to support ASME Code acceptance of the technology. This progress report covers the programmatic work from October 1985 through September 1986. 45 figs., 8 tabs.

Doctor, S.R.; Hall, T.E.; Reid, L.D.; Mart, G.A.

1987-07-01T23:59:59.000Z

242

Research Program of a Super Fast Reactor  

SciTech Connect (OSTI)

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

243

The development of self aligning geodesic dome geometries and their application at Chernobyl  

SciTech Connect (OSTI)

William M. Hardy and the author have been actively working towards the goal of establishing a non-profit engineering firm to provide remedial answers to environmental crises. Their immediate concerns are nuclear waste. Their first project has been to provide a design for the second sarcophagus planned at the Chernobyl nuclear power plant in the Ukraine. The sarcophagus will provide a habitat in which to robotically dismantle the power plant and dispose of the radioactive material over a 100+ year period. The Chernobyl Ukritiye Dome structure will `float` on a sufficiently wide, concentrically ``toothed`` footing, supported on a bed of crushed rock so that a minimum of vibrational site disturbance is achieved during construction. This will eliminate the need for a conventional poured concrete foundation and the associated excavating. The Chernobyl Ukritiye Dome will potentially be the single largest structure ever erected in the world. A goal and benefit of this project will be a minimum of exposure to radioactivity for the labor pool because of extensive off site fabrication and as much of the construction as possible will be done robotically. Not only will this structure provide superior long term encapsulation, but once completed, it will become the premiere storage location for the entire region during the next several hundred years.

Hoelzen, E.C.

1995-12-31T23:59:59.000Z

244

Lead, zinc, and strontium in limestone cap rock from Tatum salt dome, Mississippi  

SciTech Connect (OSTI)

Limestone cap rock at Tatum salt dome, Mississippi, contains disseminated pyrite, sphalerite, and galena, and disseminated to massive amounts of strontianite (SrCO/sub 3/) and celestite (SrSO/sub 4/). Sulfide minerals are locally present in bitumen-rich areas of the upper, massive portion of the limestone cap rock, whereas strontium minerals are disseminated throughout this zone. However, sulfide and strontium minerals are most abundant in the lower banded portion of the limestone cap rock, which consists of alternating subhorizontal light and dark-colored bands. The dark bands are composed of calcite of variable grain size, sulfides, quartz, dolomite, albite, and up to 1% bitumen that apparently formed by the biodegradation of crude oil. Lighter bands are composed of variable amounts of coarsely crystalline, euhedral calcite, strontianite, and celestite resulting in strontium (Sr) contents of up to 30% locally. Banded limestone cap rock at Tatum dome formed at the top of the actively dissolving anhydrite zone by a combination of sulfate reduction and oxidation of liquid hydrocarbons by bacteria to cause the precipitation of calcite and sulfide minerals and the accumulation of insoluble residue from the anhydrite (quartz, albite, dolomite). Lead and zinc in the sulfide minerals could have been derived from the dissolving anhydrite, but the abundance of Sr minerals present requires an external source. Present-day oil field brines in central Mississippi contain up to 3000 ppm Sr, and basin brines of similar composition apparently contributed Sr to the cap-rock environment during formation.

Saunders, J.A.

1988-09-01T23:59:59.000Z

245

Identification of damage in dome-like structures using hybrid sensor measurements and artificial neural networks  

Science Journals Connector (OSTI)

A damage detection scheme using multi-type sensor-based hybrid sensing and artificial-neural-network- (ANN-) based information processing was developed for dome-like structures used in civil infrastructure. Accelerometers and strain sensors were used to provide a hybrid measurement with the purpose of acquiring rich information associated with structural damage. The optimal placement of multiple sensors was explored so as to capture the most appropriate and sensitive signal features (damage parameter vectors) for damage characterization. A back-propagation ANN was constructed with the inputs extracted from the hybrid measurement. To validate the capacity of the proposed damage identification scheme, finite element analysis was conducted to identify damage in a Schwedler dome structure as an example. The performance of ANNs, trained by three kinds of damage parameter vector extracted from signals captured by (i) a sole accelerometer, (ii) a sole strain sensor, and (iii) both kinds of sensor was compared, to observe that the one trained by hybrid sensor measurement outperformed the others. Error analysis for a series of parametric studies, in which noise at different levels was included in the training input, was further carried out, and robustness of the proposed damage identification scheme under noisy measurement was demonstrated.

Wei Lu; Jun Teng; Youlin Xu; Zhongqing Su

2013-01-01T23:59:59.000Z

246

A Study and Comparison of SCR Reaction Kinetics from Reactor...  

Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

Data A Study and Comparison of SCR Reaction Kinetics from Reactor and Engine Experimental Data Presents experimental study of a Cu-zeolite SCR in both reactor and engine test cell,...

247

Sub–inertial dynamics of density–driven flows in a continuously stratified fluid on a sloping bottom. II. Isolated eddies and radiating cold domes  

Science Journals Connector (OSTI)

...eddies and radiating cold domes Francis J. Poulin Gordon E. Swaters Applied Mathematics...and radiating cold domes By Francis J. Poulin and Gordon E. Swaters Applied Mathematics...Introduction This paper is a continuation of Poulin & Swaters (1999), hereafter referred...

1999-01-01T23:59:59.000Z

248

The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases  

SciTech Connect (OSTI)

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

2012-10-01T23:59:59.000Z

249

SOFTWARE TESTING, VERIFICATION AND RELIABILITY Softw. Test. Verif. Reliab. (2014)  

E-Print Network [OSTI]

. An industrial case study using reactor protection system software shows that the automatically generated test defined by the International Electrotechnical Commission (IEC). Industrial FBD testing relies mostlySOFTWARE TESTING, VERIFICATION AND RELIABILITY Softw. Test. Verif. Reliab. (2014) Published online

250

Light Water Reactors Technology Development - Nuclear Reactors  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

251

Thermionic Reactor Design Studies  

SciTech Connect (OSTI)

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01T23:59:59.000Z

252

EIS-0021: Strategic Petroleum Reserve, Seaway Group Salt Domes, Brazoria County, Texas (also see EIS-0075-S and EIS-0029)  

Broader source: Energy.gov [DOE]

The U.S. Department of Energy's Strategic Petroleum Reserve Office developed this statement to analyze the environmental impacts which would occur during site preparation and operation of oil storage facilities at each of five proposed candidate sites in the Seaway Group of salt domes.

253

Photocatalytic reactor  

DOE Patents [OSTI]

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

1999-01-19T23:59:59.000Z

254

STATEMENT OF CONSID ERATIONS CLASS WAIVER OF THE GOVERNMENT'S DOMES'I'!C: AND FO  

Broader source: Energy.gov (indexed) [DOE]

CONSID CONSID ERATIONS CLASS WAIVER OF THE GOVERNMENT'S DOMES'I'!C: AND FO REIGN PATENT RIGHTS AND ALLOGATION OF DATA RIGHTS ARlSlNG FROM THE USE OF DOE FACI L!TI ES ANP FACILITY CON1'RACTORS BY OR FOR Tl !IRD-PARTY SPONSORS: DOE W A.IVER NO. W(C)-2011-009. Introduction The Deparlrnenr of Energy (and its predecessor agencies) (collectively. '·DOE'' or "Department") considers each of its DOE Facilities (i.e .. National Laboratori es, single-purpose research facilities, and other Department faci lities hereinafter referred to individually as '·Facility'' or collectively as "Facilities'') a unique and val uable national resource that should be made available to the extent feas ible fo r non- Federa l research and development activities and studies for third-party Sponsors.

255

PHOTOMETRY OF VARIABLE STARS FROM DOME A, ANTARCTICA: RESULTS FROM THE 2010 OBSERVING SEASON  

SciTech Connect (OSTI)

We present results from a season of observations with the Chinese Small Telescope ARray, obtained over 183 days of the 2010 Antarctic winter. We carried out high-cadence time-series aperture photometry of 9125 stars with i ?< 15.3 mag located in a 23 deg{sup 2} region centered on the south celestial pole. We identified 188 variable stars, including 67 new objects relative to our 2008 observations, thanks to broader synoptic coverage, a deeper magnitude limit, and a larger field of view. We used the photometric data set to derive site statistics from Dome A. Based on two years of observations, we find that extinction due to clouds at this site is less than 0.1 and 0.4 mag during 45% and 75% of the dark time, respectively.

Wang, Lingzhi; Zhu, Zonghong [Department of Astronomy, Beijing Normal University, Beijing 100875 (China); Macri, Lucas M.; Wang, Lifan [Mitchell Institute for Fundamental Physics and Astronomy, Department of Physics and Astronomy, Texas A and M University, College Station, TX 77843 (United States); Ashley, Michael C. B.; Lawrence, Jon S.; Luong-Van, Daniel; Storey, John W. V. [School of Physics, University of New South Wales, NSW 2052 (Australia); Cui, Xiangqun; Feng, Long-Long; Gong, Xuefei; Liu, Qiang; Shang, Zhaohui; Yang, Huigen; Yang, Ji; Yuan, Xiangyan; Zhou, Xu; Zhu, Zhenxi [Chinese Center for Antarctic Astronomy, Nanjing 210008 (China); Pennypacker, Carl R. [Center for Astrophysics, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); York, Donald G., E-mail: wanglingzhi@bao.ac.cn [Department of Astronomy and Astrophysics and Enrico Fermi Institute, University of Chicago, Chicago, IL 60637 (United States)

2013-12-01T23:59:59.000Z

256

ON THE NATURE OF RECONNECTION AT A SOLAR CORONAL NULL POINT ABOVE A SEPARATRIX DOME  

SciTech Connect (OSTI)

Three-dimensional magnetic null points are ubiquitous in the solar corona and in any generic mixed-polarity magnetic field. We consider magnetic reconnection at an isolated coronal null point whose fan field lines form a dome structure. Using analytical and computational models, we demonstrate several features of spine-fan reconnection at such a null, including the fact that substantial magnetic flux transfer from one region of field line connectivity to another can occur. The flux transfer occurs across the current sheet that forms around the null point during spine-fan reconnection, and there is no separator present. Also, flipping of magnetic field lines takes place in a manner similar to that observed in the quasi-separatrix layer or slip-running reconnection.

Pontin, D. I. [Division of Mathematics, University of Dundee, Dundee DD1 4HN (United Kingdom); Priest, E. R. [School of Mathematics and Statistics, University of St Andrews, Fife KY16 9SS (United Kingdom); Galsgaard, K., E-mail: dpontin@maths.dundee.ac.uk [Niels Bohr Institute, Copenhagen DK-2100 (Denmark)

2013-09-10T23:59:59.000Z

257

Teapot Dome: Site Characterization of a CO2- Enhanced Oil Recovery Site in Eastern Wyoming  

SciTech Connect (OSTI)

Naval Petroleum Reserve No. 3 (NPR-3), better known as the Teapot Dome oil field, is the last U.S. federally-owned and -operated oil field. This provides a unique opportunity for experiments to provide scientific and technical insight into CO{sub 2}-enhanced oil recovery (EOR) and other topics involving subsurface fluid behavior. Towards that end, a combination of federal, academic, and industrial support has produced outstanding characterizations of important oil- and brine-bearing reservoirs there. This effort provides an unparalleled opportunity for industry and others to use the site. Data sets include geological, geophysical, geochemical, geomechanical, and operational data over a wide range of geological boundary conditions. Importantly, these data, many in digital form, are available in the public domain due to NPR-3's federal status. Many institutions are already using portions of the Teapot Dome data set as the basis for a variety of geoscience, modeling, and other research efforts. Fifteen units, 9 oil-bearing and 6 brine-bearing, have been studied to varying degrees. Over 1200 wells in the field are active or accessible, and over 400 of these penetrate 11 formations located below the depth that corresponds to the supercritical point for CO{sub 2}. Studies include siliciclastic and carbonate reservoirs; shale, carbonate, and anhydrite cap rocks; fractured and unfractured units; and over-pressured and under-pressured zones. Geophysical data include 3D seismic and vertical seismic profiles. Reservoir data include stratigraphic, sedimentological, petrologic, petrographic, porosity, and permeability data. These have served as the basis for preliminary 3D flow simulations. Geomechanical data include fractures (natural and drilling induced), in-situ stress determination, pressure, and production history. Geochemical data include soil gas, noble gas, organic, and other measures. The conditions of these reservoirs directly or indirectly represent many reservoirs in the U.S., Canada, and overseas.

Friedmann, S J; Stamp, V

2005-11-01T23:59:59.000Z

258

Intrapleural Fluid Infusion for MR-Guided High-Intensity Focused Ultrasound Ablation in the Liver Dome  

Science Journals Connector (OSTI)

Rationale and Objectives Magnetic resonance–guided high-intensity focused ultrasound (MR-HIFU) ablation of tumors in the liver dome is challenging because of the presence of air in the costophrenic angle. In this study, we used a porcine liver model and a clinical MR-HIFU system to assess the feasibility and safety of using intrapleural fluid infusion (IPI) to create an acoustic window for MR-HIFU ablation in the liver dome. Materials and Methods Healthy adult Dalland land pigs (n = 6) under general anesthesia were used with animal committee approval. Degassed saline (200–800 mL) was infused into the intrapleural space under ultrasound guidance. A clinical 1.5-T MR-HIFU system was used to perform sonications (4-mm treatment cells, 300–450 W, 20–30 seconds) in the liver dome under real-time MR thermometry. An intercostal firing technique was used to prevent rib heating in one experiment. Technical success was defined as a temperature increase (>10°C) in the target area. After termination, the animal was examined for thermal damage to liver, diaphragm, pleura, lung, or intercostal muscle. Results An acoustic window was established in all animals. A temperature increase in the target area was achieved in all animals (max. 47°C–67°C). MR thermometry showed no heating outside the target area. Intercostal firing effectively reduced rib heating (55°C vs. 42°C). Postmortem examination revealed no unwanted thermal damage. One complication occurred, in the first experiment, because of an ill-suited needle (displacement of the needle). Conclusions The results indicate that IPI may be used safely to assist MR-HIFU ablation of tumors in the liver dome. For reliable tissue coagulation, IPI must be combined with an intercostal sonication technique. Considering the proportion of patients with tumors in the liver dome, IPI widens the applicability of MR-HIFU ablation for liver tumors considerably.

Joost W. Wijlemans; Martijn de Greef; Gerald Schubert; Chrit T.W. Moonen; Maurice A.A.J. van den Bosch; Mario Ries

2014-01-01T23:59:59.000Z

259

Fluid flow in the resurgent dome of Long Valley Caldera: implications from thermal data and deep electrical sounding  

Science Journals Connector (OSTI)

Temperatures of 100°C are measured at 3 km depth in a well located on the resurgent dome in the center of Long Valley Caldera, California, despite an assumed >800°C magma chamber at 6–8 km depth. Local downflow of cold meteoric water as a process for cooling the resurgent dome is ruled out by a Peclét-number analysis of temperature logs. These analyses reveal zones with fluid circulation at the upper and lower boundaries of the Bishop Tuff, and an upflow zone in the metasedimentary rocks. Vertical Darcy velocities range from 10 to 70 cm a?1. A 21-km-long geoelectrical profile across the caldera provides resistivity values to the order of 100 to >103 ?m down to a depth of 6 km, as well as variations of self-potential. Interpretation of the electrical data with respect to hydrothermal fluid movement confirms that there is no downflow beneath the resurgent dome. To explain the unexpectedly low temperatures in the resurgent dome, we challenge the common view that the caldera as a whole is a regime of high temperatures and the resurgent dome is a local cold anomaly. Instead, we suggest that the caldera was cooled to normal thermal conditions by vigorous hydrothermal activity in the past, and that a present-day hot water flow system is responsible for local hot anomalies, such as Hot Creek and the area of the Casa Diablo geothermal power plant. The source of hot water has been associated with recent shallow intrusions into the West Moat. The focus of planning for future power plants should be to locate this present-day flow system instead of relying on heat from the old magma chamber.

Daniel F.C Pribnow; Claudia Schütze; Suzanne J Hurter; Christina Flechsig; John H Sass

2003-01-01T23:59:59.000Z

260

Safety evaluation report related to the Department of Energy`s proposal for the irradiation of lead test assemblies containing tritium-producing burnable absorber rods in commercial light-water reactors. Project Number 697  

SciTech Connect (OSTI)

The NRC staff has reviewed a report, submitted by DOE to determine whether the use of a commercial light-water reactor (CLWR) to irradiate a limited number of tritium-producing burnable absorber rods (TPBARs) in lead test assemblies (LTAs) raises generic issues involving an unreviewed safety question. The staff has prepared this safety evaluation to address the acceptability of these LTAs in accordance with the provision of 10 CFR 50.59 without NRC licensing action. As summarized in Section 10 of this safety evaluation, the staff has identified issues that require NRC review. The staff has also identified a number of areas in which an individual licensee undertaking irradiation of TPBAR LTAs will have to supplement the information in the DOE report before the staff can determine whether the proposed irradiation is acceptable at a particular facility. The staff concludes that a licensee undertaking irradiation of TPBAR LTAs in a CLWR will have to submit an application for amendment to its facility operating license before inserting the LTAs into the reactor.

NONE

1997-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Hybrid adsorptive membrane reactor  

DOE Patents [OSTI]

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01T23:59:59.000Z

262

In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements  

SciTech Connect (OSTI)

Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

2012-09-17T23:59:59.000Z

263

GEN-IV Reactors  

Science Journals Connector (OSTI)

Generation-IV reactors are a set of nuclear reactors currently being developed under international collaborations targeting ... economics, proliferation resistance, and physical protection of nuclear energy. Nuclear

Taek K. Kim

2013-01-01T23:59:59.000Z

264

The Netherlands Reactor Centre  

Science Journals Connector (OSTI)

... Two illustrated brochures in English have recently J. been issued by the Netherlands Reactor Centre ( ... Centre (Reactor Centrum Nederland). The first* gives a general survey of the ...

S. WEINTROUB

1964-04-04T23:59:59.000Z

265

Nuclear Reactor Safety Design Criteria  

Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Certified 11-18-10.

1993-01-19T23:59:59.000Z

266

Overview of Sandia National Laboratories pulse nuclear reactors  

SciTech Connect (OSTI)

Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.

Schmidt, T.R. [Sandia National Labs., Albuquerque, NM (United States); Reuscher, J.A. [Texas A& M Univ., College Station, TX (United States)

1994-10-01T23:59:59.000Z

267

Consortium for Advanced Simulation of Light Water Reactors (CASL...  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

has designed and operated 52 test reactors, including EBR-1, the world's first nuclear power plant Key Contributions System safety analysis Multiscale fuel performance...

268

Gas-cooled fast breeder reactor. Quarterly progress report, November 1, 1979 through January 31, 1980  

SciTech Connect (OSTI)

Information is presented concerning the nuclear steam supply system; reactor core; systems engineering; safety and reliability; and circulator test facility.

Not Available

1980-02-01T23:59:59.000Z

269

Rocky Mountain Oilfield Testing Center RMOTC at the Naval Petroleum Reserve No. 3  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

OILFIELD TESTING OILFIELD TESTING OILFIELD TESTING CENTER CENTER 2 2 HISTORY OF TEAPOT DOME Mark Milliken 3 3 TEAPOT DOME LOCATION 4 4 Salt Creek 670 MMBBLS 722 BCF Teapot Dome 27 MMBBLS 57 BCF N P R - 3 Cumulative Production 5 5 The Great White Fleet December 1907 - February 1909 6 6 THE END OF COAL-FIRED SHIPS * 2-week cruising time. * Labor and time intensive cleaning and reloading. * At the mercy of foreign countries for coal supply. * 1912: All battleships will be oil-powered. Great White Fleet 7 7 NAVAL PETROLEUM RESERVES ARE BORN * 1908: Dr. Otis Smith, USGS Director, recommends DOI retain oil lands for fuel reserve for Navy. * 1909: Taft withdraws 3,000,000 acres in Wyoming and California. * 1910: Concern over the President's authority to withdraw lands, so Congress passed the Pickett Act. * 1910: Taft issues a 2nd land withdrawal executive

270

SRS Small Modular Reactors  

SciTech Connect (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2012-04-27T23:59:59.000Z

271

SRS Small Modular Reactors  

ScienceCinema (OSTI)

The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

None

2014-05-21T23:59:59.000Z

272

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect (OSTI)

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20T23:59:59.000Z

273

EIS-0024: Strategic Petroleum Reserve, Capline Group Salt Domes, Iberia, Napoleonville, Weeks Island Expansion, Bayou Choctaw Expansion, Chacahoula- Iberia, Iberville, and Lafourche Parishes, Louisiana  

Broader source: Energy.gov [DOE]

The Strategic Petroleum Reserves developed this EIS to analyze the environmental impacts which would occur during site preparation and operation of oil storage facilities at each of five proposed candidate sites in the Capline Group of salt domes.

274

Experience in testing of a solution mined storage cavern. [Strategic Petroleum Reserve  

SciTech Connect (OSTI)

Recertification tests were made of the Department of Energy/Strategic Petroleum Reserve oil storage cavern number 6 in the West Hackberry, Louisiana Salt Dome. The cavern has a volume of 8,600,000 barrels. Tests included hydrostatic tests of the brine filled cavern and nitrogen leak tests of the three wells entering the cavern. Test procedures are described and test results are discussed.

Goin, K.L.

1982-01-01T23:59:59.000Z

275

Nuclear reactor  

DOE Patents [OSTI]

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16T23:59:59.000Z

276

Tatum Dome field study report and monitoring data analysis: A supplemental report  

SciTech Connect (OSTI)

This report is a supplement to the Water Resources Center, Desert Research Institute report, DOE/NV/10384-03, ``Tatum Dome Field Study Report and Monitoring Data Analysis,`` Water Resources Center Publication No. 45044. The field study was initiated during the Spring of 1984 because of persistent tritium concentrations in the Surficial Aquifer determined from observed annual water samples from the series of hydrologic monitoring holes (HMH). An anomalous increase in tritium concentrations in monthly water samples from some of the hydrologic monitoring holes was also observed during the Spring of 1984 by the State of Mississippi, Division of Radiological Health. This Spring increase in tritium concentrations may well have been present earlier, but was not recognized because monthly tritium concentration data were not collected prior to 1984. It is hypothesized that groundwater in the Surficial Aquifer is made up of two layers. The older and deeper water within the Surficial Aquifer contains tritium contamination. The shallower water, infiltrating from recent precipitation, is essentially tritium free. These waters do not naturally mix completely and are only significantly mixed in the hydrologic monitoring holes by the sampling procedure. The quantity of shallow infiltrating precipitation available for mixing varies inversely with the rate of evapotranspiration. Since dispersive mixing along the boundary between the two waters also occurs as a result of groundwater movement, the average concentration of tritium in the Surficial Aquifer is decreased by dilution as well as radioactive decay.

Fenske, P.R.

1989-09-01T23:59:59.000Z

277

Tatum Dome field study report and monitoring data analysis: A supplemental report  

SciTech Connect (OSTI)

This report is a supplement to the Water Resources Center, Desert Research Institute report, DOE/NV/10384-03, Tatum Dome Field Study Report and Monitoring Data Analysis,'' Water Resources Center Publication No. 45044. The field study was initiated during the Spring of 1984 because of persistent tritium concentrations in the Surficial Aquifer determined from observed annual water samples from the series of hydrologic monitoring holes (HMH). An anomalous increase in tritium concentrations in monthly water samples from some of the hydrologic monitoring holes was also observed during the Spring of 1984 by the State of Mississippi, Division of Radiological Health. This Spring increase in tritium concentrations may well have been present earlier, but was not recognized because monthly tritium concentration data were not collected prior to 1984. It is hypothesized that groundwater in the Surficial Aquifer is made up of two layers. The older and deeper water within the Surficial Aquifer contains tritium contamination. The shallower water, infiltrating from recent precipitation, is essentially tritium free. These waters do not naturally mix completely and are only significantly mixed in the hydrologic monitoring holes by the sampling procedure. The quantity of shallow infiltrating precipitation available for mixing varies inversely with the rate of evapotranspiration. Since dispersive mixing along the boundary between the two waters also occurs as a result of groundwater movement, the average concentration of tritium in the Surficial Aquifer is decreased by dilution as well as radioactive decay.

Fenske, P.R.

1989-09-01T23:59:59.000Z

278

Characterization of cement from a well at Teapot Dome Oil Field: Implications for geological sequestration  

Science Journals Connector (OSTI)

Wellbores represent the weakest link in terms of CO2 storage permanence. As a result, special attention to the numerous existing wells that perforate storage formations is needed. The pre-injection condition of the cement can influence the rate (and type) of alteration by the injected CO2 plume. The condition of the existing well cement depends on a variety of factors including wellbore/formation and wellbore/brine interactions as well as the composition and type of cement placed in the well (i.e. type of admixtures used, water/solids ratio, sulfate resistant mixes, etc.). In this paper, the details of recovering wellbore cement from an older well to determine pre-injection seal integrity are described. Petrographical and chemical analyses are presented for samples of cement that were retrieved from a 19-year-old well at Teapot Dome in Wyoming. Examination revealed that the retrieved cement had altered as a result of original slurry composition and with respect to the local downhole wellbore environment. Although samples were obtained from a single well, significant differences were observed in their alteration and condition. Sulfate attack resulted in abundant ettringite formation in a cement sample taken adjacent to the Wall Creek sandstone (3060 ft), while cement taken adjacent to the Tensleep formation (5478 ft) was decalcified and enriched in magnesium, owing to reaction of calcium hydroxide in the cement with the dolomitic formation.

George W. Scherer; Barbara Kutchko; Niels Thaulow; Andrew Duguid; Bryant Mook

2011-01-01T23:59:59.000Z

279

Questions and Answers - Why are the Halls in bio-dome shapes?  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

How much does it cost a yearto run Jefferson Lab? How much does it cost a year<br>to run Jefferson Lab? Previous Question (How much does it cost a year to run Jefferson Lab?) Questions and Answers Main Index Next Question (Why did it take so long to build Jefferson Lab?) Why did it take so longto build Jefferson Lab? Why are the Halls in bio-dome shapes? The answer to your question is the answer to many questions... money. The shape of our experimental halls was that which could do the job and spend the least amount of money. There are several reasons for this that you won't quite understand unless you have taken geometry. If you need to enclose a certain amount of AREA, but have to pay for the LENGTH of wall you use, you want to build whatever type of building will enclose that AREA with the shortest LENGTH of wall. It just so happens a circle encloses the

280

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility  

SciTech Connect (OSTI)

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

1994-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

SECONDARY NATURAL GAS RECOVERY IN THE APPALACHIAN BASIN: APPLICATION OF ADVANCED TECHNOLOGIES IN A FIELD DEMONSTRATION SITE, HENDERSON DOME, WESTERN PENNSYLVANIA  

SciTech Connect (OSTI)

The principal objectives of this project were to test and evaluate technologies that would result in improved characterization of fractured natural-gas reservoirs in the Appalachian Basin. The Bureau of Economic Geology (Bureau) worked jointly with industry partner Atlas Resources, Inc. to design, execute, and evaluate several experimental tests toward this end. The experimental tests were of two types: (1) tests leading to a low-cost methodology whereby small-scale microfractures observed in matrix grains of sidewall cores can be used to deduce critical properties of large-scale fractures that control natural-gas production and (2) tests that verify methods whereby robust seismic shear (S) waves can be generated to detect and map fractured reservoir facies. The grain-scale microfracture approach to characterizing rock facies was developed in an ongoing Bureau research program that started before this Appalachian Basin study began. However, the method had not been tested in a wide variety of fracture systems, and the tectonic setting of rocks in the Appalachian Basin composed an ideal laboratory for perfecting the methodology. As a result of this Appalachian study, a low-cost commercial procedure now exists that will allow Appalachian operators to use scanning electron microscope (SEM) images of thin sections extracted from oriented sidewall cores to infer the spatial orientation, relative geologic timing, and population density of large-scale fracture systems in reservoir sandstones. These attributes are difficult to assess using conventional techniques. In the Henderson Dome area, large quartz-lined regional fractures having N20E strikes, and a subsidiary set of fractures having N70W strikes, are prevalent. An innovative method was also developed for obtaining the stratigraphic and geographic tops of sidewall cores. With currently deployed sidewall coring devices, no markings from which top orientation can be obtained are made on the sidewall core itself during drilling. The method developed in this study involves analysis of the surface morphology of the broken end of the core as a top indicator. Together with information on the working of the tool (rotation direction), fracture-surface features, such as arrest lines and plume structures, not only give a top direction for the cores but also indicate the direction of fracture propagation in the tough, fine-grained Cataract/Medina sandstones. The study determined that microresistivity logs or other image logs can be used to obtain accurate sidewall core azimuths and to determine the precise depths of the sidewall cores. Two seismic S-wave technologies were developed in this study. The first was a special explosive package that, when detonated in a conventional seismic shot hole, produces more robust S-waves than do standard seismic explosives. The importance of this source development is that it allows S-wave seismic data to be generated across all of the Appalachian Basin. Previously, Appalachian operators have not been able to use S-wave seismic technology to detect fractured reservoirs because the industry-standard S-wave energy source, the horizontal vibrator, is not a practical source option in the heavy timber cover that extends across most of the basin. The second S-wave seismic technology that was investigated was used to verify that standard P-wave seismic sources can create robust downgoing S-waves by P-to-S mode conversion in the shallow stratigraphic layering in the Appalachian Basin. This verification was done by recording and analyzing a 3-component vertical seismic profile (VSP) in the Atlas Montgomery No. 4 well at Henderson Dome, Mercer County, Pennsylvania. The VSP data confirmed that robust S-waves are generated by P-to-S mode conversion at the basinwide Onondaga stratigraphic level. Appalachian operators can thus use converted-mode seismic technology to create S-wave images of fractured and unfractured rock systems throughout the basin.

BOB A. HARDAGE; ELOISE DOHERTY; STEPHEN E. LAUBACH; TUCKER F. HENTZ

1998-08-14T23:59:59.000Z

282

Attrition reactor system  

DOE Patents [OSTI]

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Scott, C.D.; Davison, B.H.

1993-09-28T23:59:59.000Z

283

Elementary Reactor Physics  

Science Journals Connector (OSTI)

... THERE are few subjects which have developed at the rate at which reactor physics and ... physics and reactor theory have done. This, of course, is largely due to the circumstances in ...

J. F. HILL

1962-02-10T23:59:59.000Z

284

Colliding Beam Fusion Reactors  

Science Journals Connector (OSTI)

The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker–Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are ... the rec...

Norman Rostoker; Artan Qerushi; Michl Binderbauer

2003-06-01T23:59:59.000Z

285

Parallel Monte Carlo reactor neutronics  

SciTech Connect (OSTI)

The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved.

Blomquist, R.N.; Brown, F.B.

1994-03-01T23:59:59.000Z

286

Prospects for spheromak fusion reactors  

Science Journals Connector (OSTI)

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on...

T. K. Fowler; D. D. Hua

1995-06-01T23:59:59.000Z

287

RESULTS OF FIELD TESTING DOE  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

of Energy facility located within Naval Petroleum Reserve No. 3 (NPR-3), also known as Teapot Dome Oil Field, about 35 miles north of Casper, Wyoming. Teapot Dome Field, Natrona...

288

Improved vortex reactor system  

DOE Patents [OSTI]

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

1995-01-01T23:59:59.000Z

289

The Argonaut Reactor - Reactors designed/built by Argonne National  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy

290

High Temperature Test Laboratory Accomplishments  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Knudson, K. Condie, and B. Sencer, "In-situ Creep Testing Capability for the Advanced Test Reactor," Nuclear Technology, 179, 3, September 2012, pp 413-428. B. Geslot, T. Unruh,...

291

Mineralogic and isotopic constraints on the origin of strontium-rich cap rock, Tatum Dome, Mississippi, U.S.A.  

Science Journals Connector (OSTI)

The limestone portion of the salt dome cap rock at Tatum dome, Mississippi, is composed of an upper massive zone and a lower banded zone. The upper zone consists of equigranular fine-crystalline calcite with veinlets and disseminations of carbonaceous matter associated with minor amounts of detrital and authigenic quartz, sulfides and Sr minerals. The lower zone is composed of alternating light and dark calcite bands. The dark bands are composed of fine-grained and peloidal calcite, quartz, bitumen and disseminated sulfide minerals. The lighter bands consist of variable proportions of generally coarse-crystalline euhedral calcite, celestite and strontianite resulting in Sr contents of up to 30% locally. Solubility data for celestite, strontianite, calcite and anhydrite suggest that a decrease in temperature favors the replacement of Ca minerals by Sr minerals, which is consistent with the observed mineral textures and paragenesis. However, the source of cap rock Sr is difficult to determine. Anhydrite at Tatum dome contains 800 ppm Sr, but the abundance of Sr minerals in the limestone cap rock and the 87Sr86Sr ratios of limestone cap rock minerals require a Sr source other than local anhydrite. Sr released by the replacement of anhydrite by calcite is capable of producing a molar ratio of celestite to calcite of only ? 0.001, yet locally this ratio is ? 3. A likely source of additional Sr is oilfield brines, such as those in central Mississippi that contain up to 3000 ppm Sr along with significant Pb and Zn. Episodic introduction of brines into the cap rock in combination with the action of sulfate-reducing bacteria probably caused the sequential production of the dark bands. The lighter bands probably reflect introduction of a later Sr-enriched fluid or evolution of the original fluid in combination with changes in the chemical parameters controlling mineral precipitation. Calcite was replaced by Sr minerals during this later Sr-rich event.

James A. Saunders; James D. Prikryl; Harry H. Posey

1988-01-01T23:59:59.000Z

292

Reactor vessel support system  

DOE Patents [OSTI]

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

1982-01-01T23:59:59.000Z

293

Testing the scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test  

E-Print Network [OSTI]

in the Tokamak Fusion Test Reactor dimensionless scaling experiments D. R. Mikkelsen, S. D. Scott Princeton the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] nondimensional to International Tokamak Experimental Reactor [2] (ITER) class tokamaks. This paper compares the predictions

294

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01T23:59:59.000Z

295

Reactor water cleanup system  

DOE Patents [OSTI]

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

Gluntz, D.M.; Taft, W.E.

1994-12-20T23:59:59.000Z

296

A Study and Comparison of SCR Reaction Kinetics from Reactor and Engine Experimental Data  

Broader source: Energy.gov [DOE]

Presents experimental study of a Cu-zeolite SCR in both reactor and engine test cell, and comparison of the model parameters between the SCR reactor and engine model

297

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

298

Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application  

E-Print Network [OSTI]

HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System... to be at right angles to each other, ignoring an angular distribution of radiant heat.7 MORECA, used by ORNL, simulates accident scenarios for certain gas-cooled reactor types.7 INET conducts their analysis using Thermix, which performs two...

Moore, Eugene James Thomas

2006-08-16T23:59:59.000Z

299

Sandia National Laboratories: Reactor Component Testing  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Doppler Velocimeter EC Top Publications A Comparison of Platform Options for Deep-water Floating Offshore Vertical Axis Wind Turbines: An Initial Study Nonlinear Time-Domain...

300

Microsoft Word - ROCKY MOUNTAIN OILFIELD TESTING CENTER - STWA-AOT-10192011 -R2  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Energy Energy STWA : Viscosity Reduction Test An assessment of an in-line viscosity reduction device Naval Petroleum Reserve No. 3, Teapot Dome Field, Wyoming Final Report for October 19, 2011 This document may contain protected/confi dential information produced under and Funds-In Agreement (FIA) and is not to be further disclosed except as expressly provided for in the FIA.

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Oilfield testing center aids industry in evaluating cutting-edge innovations  

SciTech Connect (OSTI)

The Rocky Mountain Oilfield Testing Center at Teapot Dome keeps a producing field open for research and development. Using a producing oil field for research is the surest way to determine the success or failure of a new invention or technique. The field has 600 producing wells and 68 injection wells.

Duey, R.

1996-01-01T23:59:59.000Z

302

Generating Unstructured Nuclear Reactor Core Meshes in Parallel  

Science Journals Connector (OSTI)

Abstract Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

Rajeev Jain; Timothy J. Tautges

2014-01-01T23:59:59.000Z

303

Generation -IV Reactor Concepts  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),

304

Power Reactor Progress  

Science Journals Connector (OSTI)

Argonne kicks off EBWR; Allis-Chalmers plans power reactor using both nuclear and conventional fuels ... NUCLEAR POWER took two giant steps last week. ... Just as the first nuclear power system in the U. S. designed and built solely for the generation of electric power went into full operation at Argonne, Allis-Chalmers came up with a new twist in power reactors—a controlled recirculation boiling reactor (CRBR) using both nuclear and conventional fuels (C&EN, Feb. 18, page 7). ...

1957-02-25T23:59:59.000Z

305

Improved vortex reactor system  

DOE Patents [OSTI]

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Diebold, J.P.; Scahill, J.W.

1995-05-09T23:59:59.000Z

306

Assessing low power reactivity levels in subcritical CANDU reactors  

SciTech Connect (OSTI)

A new technique has been developed for monitoring slow reactivity changes in Canadian deuterium uranium (CANDU) nuclear reactors during low power operation, following a sustained period of high power operation. The power doubling halving worth (PDHW) test tracks slow reactivity changes by evaluating the reactivity perturbation required by the power regulating system to halve and double reactor power over time. During low power operation of a CANDU reactor, the PDHW test is used to monitor the decay of the photoneutron precursors, so that the reactor power can be lowered using a preset amount of reactivity. The PDHW test is described in this paper and is validated by using computer simulations and operating data from a CANDU reactor.

Teare, S.W. [Corporation of the City of Mississauga, Ontario (Canada)] [Corporation of the City of Mississauga, Ontario (Canada); Shanes, F.C. [Ontario Hydro, Toronto, Ontario (Canada)] [Ontario Hydro, Toronto, Ontario (Canada); Hersey, M.W. [Pickering Nuclear Generating Station, Ontario (Canada)] [Pickering Nuclear Generating Station, Ontario (Canada)

1996-08-01T23:59:59.000Z

307

AEC Pushes Fusion Reactors  

Science Journals Connector (OSTI)

AEC Pushes Fusion Reactors ... Project Sherwood, as the study program is called, began in 1951-52 soon after the first successful thermonuclear explosion in the Pacific. ...

1955-10-10T23:59:59.000Z

308

Tokamak reactor first wall  

DOE Patents [OSTI]

This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

1984-11-20T23:59:59.000Z

309

Advanced Nuclear Research Reactor  

SciTech Connect (OSTI)

This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.

Lolich, J.V.

2004-10-06T23:59:59.000Z

310

CASL Test Stands . . . Enabling the CASL mission  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Test Stands . . . Enabling the CASL mission In a nuclear reactor core there are many phenomena in motion at both the microscopic and macroscopic levels that ultimately affect the...

311

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...  

Office of Scientific and Technical Information (OSTI)

A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor Design and Feasibility Problem Re-direct Destination: Temp Data Fields Rosen, M. A.; Coburn, D. B.; Flynn, T....

312

Portfolio for fast reactor collaboration  

SciTech Connect (OSTI)

The development of the LMFBR type reactor in the United Kingdom is reviewed. Design characteristics of a commercial demonstration fast reactor are presented and compared with the Super Phenix reactor.

Rippon, S.

1981-12-01T23:59:59.000Z

313

Handbook of Reactor Physics  

Science Journals Connector (OSTI)

... THIS handbook is one volume in a series sponsored by the United States Atomic Energy Commission with ... data and reference information in the field of reactors. The volume is devoted to reactor physics and radiation shielding, the latter subject occupying approximately a quarter of the book.

PETER W. MUMMERY

1956-08-25T23:59:59.000Z

314

Fast reactor safety  

Science Journals Connector (OSTI)

... SIR, - In his article on fast reactor safety (26 July, page 270) Norman Dombey claims to introduce to non-specialists ... , page 270) Norman Dombey claims to introduce to non-specialists some features of fast reactors that are not available outside the technical literature. The non-specialist would do well ...

R.D. SMITH

1979-08-23T23:59:59.000Z

315

Instrumentation of Nuclear Reactors  

Science Journals Connector (OSTI)

... s Lecture Theatre on January 8, a symposium of papers on the instrumentation of nuclear reactors was organized, at which about five hundred members and visitors attended, including guests from ... the Institution, took the chair and introduced Sir John Cockcroft, whose lecture on "Nuclear Reactors and their Applications" provided a general background for the three specialized papers which followed. ...

1953-03-07T23:59:59.000Z

316

Nuclear Research Reactors  

Science Journals Connector (OSTI)

... their countries for the advent of nuclear power. A few countries had built large research reactors for the production of isotopes and to study the behaviour of nuclear fuel, but ... production of isotopes and to study the behaviour of nuclear fuel, but the small training reactor had not been developed. Since then, research ...

T. E. ALLIBONE

1963-07-20T23:59:59.000Z

317

The integral fast reactor fuel cycle  

SciTech Connect (OSTI)

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01T23:59:59.000Z

318

Journal of Volcanology and Geothermal Research69 (1995) 105-l 16 Mount St. Helens and Santiaguito lava domes: The effect of short-  

E-Print Network [OSTI]

of silicic lava flows, we studied surface characteristics and obtained water content and hydrogen isotopic for these patterns is most clearly preserved in lavas erupted during early, rapid stages of dome growth. Petrologists to sample flows early in their emplacement while paying attention to surface texture, position relative

Rose, William I.

319

Growth of Dome-Shaped Carbon Nanoislands on Ir(111): The Intermediate between Carbidic Clusters and Quasi-Free-Standing Graphene  

E-Print Network [OSTI]

of hydrocarbon dissociation on transition metal (TM) sur- faces represents a challenging way to its synthesisGrowth of Dome-Shaped Carbon Nanoislands on Ir(111): The Intermediate between Carbidic Clusters coupled carbidic carbon and a quasi-free-standing graphene layer, can provide information for a rational

Alfè, Dario

320

Canadian university research reactors  

SciTech Connect (OSTI)

In Canada there are seven university research reactors: one medium-power (2-MW) swimming pool reactor at McMaster University and six low-power (20-kW) SLOWPOKE reactors at Dalhousie University, Ecole Polytechnique, the Royal Military College, the University of Toronto, the University of Saskatchewan, and the University of Alberta. This paper describes primarily the McMaster Nuclear Reactor (MNR), which operates on a wider scale than the SLOWPOKE reactors. The MNR has over a hundred user groups and is a very broad-based tool. The main applications are in the following areas: (1) neutron activation analysis (NAA); (2) isotope production; (3) neutron beam research; (4) nuclear engineering; (5) neutron radiography; and (6) nuclear physics.

Ernst, P.C.; Collins, M.F.

1989-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Reactor & Nuclear Systems Publications | ORNL  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Nuclear Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications...

322

A method of estimating time scales of atmospheric piston and its application at DomeC (Antarctica)  

E-Print Network [OSTI]

Temporal fluctuations of the atmospheric piston are critical for interferometers as they determine their sensitivity. We characterize an instrumental set-up, termed the piston scope, that aims at measuring the atmospheric time constant, tau0, through the image motion in the focal plane of a Fizeau interferometer. High-resolution piston scope measurements have been obtained at the observatory of Paranal, Chile, in April 2006. The derived atmospheric parameters are shown to be consistent with data from the astronomical site monitor, provided that the atmospheric turbulence is displaced along a single direction. Piston scope measurements, of lower temporal and spatial resolution, were for the first time recorded in February 2005 at the Antarctic site of DomeC. Their re-analysis in terms of the new data calibration sharpens the conclusions of a first qualitative examination.

A. Kellerer; M. Sarazin; T. Butterley; R. Wilson

2007-03-06T23:59:59.000Z

323

The CANDU Reactor System: An Appropriate Technology  

Science Journals Connector (OSTI)

...apparent. The devel-opment of Zircaloy-2 by the U.S. Bettis Laboratory in Pittsburgh, made this con-cept practicable...necessary to base a reactor de-sign on Zircaloy-2 had not Bettis work-ers been testing the material in NRX as part of a collaborative...

J. A. L. Robertson

1978-02-10T23:59:59.000Z

324

Creation of a neutrino laboratory for search for sterile neutrino at SM-3 reactor  

E-Print Network [OSTI]

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Serebrov, A P; Samoylov, R M; Fomin, A K; Zinoviev, V G; Neustroev, P V; Golovtsov, V L; Gruzinsky, N V; Solovey, V A; Cherniy, A V; Zherebtsov, O M; Martemyanov, V P; Zinoev, V G; Tarasenkov, V G; Aleshin, V I; Petelin, A L; Pavlov, S V; Izhutov, A L; Sazontov, S A; Ryazanov, D K; Gromov, M O; Afanasiev, V V; Matrosov, L N; Matrosova, M Yu

2015-01-01T23:59:59.000Z

325

Light Water Reactors [Corrosion and Mechanics of Materials] - Nuclear  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Light Water Reactors Light Water Reactors Capabilities Materials Testing Environmentally Assisted Cracking (EAC) of Reactor Materials Corrosion Performance/Metal Dusting Overview Light Water Reactors Fatigue Testing of Carbon Steels and Low-Alloy Steels Environmentally Assisted Cracking of Ni-Base Alloys Irradiation-Induced Stress Corrosion Cracking of Austenitic Stainless Steels Steam Generator Tube Integrity Program Air Oxidation Kinetics for Zr-based Alloys Fossil Energy Fusion Energy Metal Dusting Publications List Irradiated Materials Steam Generator Tube Integrity Other Facilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Corrosion and Mechanics of Materials Light Water Reactors Bookmark and Share To continue safe operation of current LWRs, the aging degradation of the

326

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect (OSTI)

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

327

Nuclear reactor control column  

DOE Patents [OSTI]

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

Bachovchin, Dennis M. (Plum Borough, PA)

1982-01-01T23:59:59.000Z

328

Transient thermal analysis of a space reactor power system  

E-Print Network [OSTI]

Thermoelectric Power Conversion Module Heat Pipe Radiator Module . Auxiliary Modules . Flow of Calculation . Transient Test Cases Studied Summary . 10 10 CHAPTER II. ENERGY EQUATION FINITE DIFFERENCING . . 12 Energy Equation for a Solid Finite..., but this stud~ uses a generic liquid metal cooled fast reactor concept as the model to test the code. The space power svstem to be modeled consists of a liquid lithium cooled fast reactor, primarv and secondary loops svith a sell-induced thermoelectric...

Gaeta, Michael J.

1988-01-01T23:59:59.000Z

329

Testing the ae \\Lambda scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test  

E-Print Network [OSTI]

in the Tokamak Fusion Test Reactor dimensionless scaling experiments D. R. Mikkelsen, S. D. Scott Princeton the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] nondimensional to extrapo� late [1] from current experiments to International Tokamak Experimental Reactor [2] (ITER) class

330

Nuclear reactor reflector  

DOE Patents [OSTI]

A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

1994-01-01T23:59:59.000Z

331

Spherical torus fusion reactor  

DOE Patents [OSTI]

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Martin Peng, Y.K.M.

1985-10-03T23:59:59.000Z

332

Solar Thermal Reactor Materials Characterization  

SciTech Connect (OSTI)

Current research into hydrogen production through high temperature metal oxide water splitting cycles has created a need for robust high temperature materials. Such cycles are further enhanced by the use of concentrated solar energy as a power source. However, samples subjected to concentrated solar radiation exhibited lifetimes much shorter than expected. Characterization of the power and flux distributions representative of the High Flux Solar Furnace(HFSF) at the National Renewable Energy Laboratory(NREL) were compared to ray trace modeling of the facility. In addition, samples of candidate reactor materials were thermally cycled at the HFSF and tensile failure testing was performed to quantify material degradation. Thermal cycling tests have been completed on super alloy Haynes 214 samples and results indicate that maximum temperature plays a significant role in reduction of strength. The number of cycles was too small to establish long term failure trends for this material due to the high ductility of the material.

Lichty, P. R.; Scott, A. M.; Perkins, C. M.; Bingham, C.; Weimer, A. W.

2008-03-01T23:59:59.000Z

333

Reactor Thermal-Hydraulics  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

334

Reactor hot spot analysis  

SciTech Connect (OSTI)

The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

Vilim, R.B.

1985-08-01T23:59:59.000Z

335

Molten metal reactors  

DOE Patents [OSTI]

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05T23:59:59.000Z

336

Testing WIMS-D4M cross sections and the ANL ENDF/B-V 69 group library. Results from global diffusion and Monte Carlo calculations compared with measurements in the Romanian 14-MW TRIGA reactor  

SciTech Connect (OSTI)

The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.

Bretscher, M.M.

1993-12-31T23:59:59.000Z

337

F Reactor Inspection  

ScienceCinema (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-11-24T23:59:59.000Z

338

F Reactor Inspection  

SciTech Connect (OSTI)

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29T23:59:59.000Z

339

Test Automation Test Automation  

E-Print Network [OSTI]

Test Automation Test Automation Mohammad Mousavi Eindhoven University of Technology, The Netherlands Software Testing 2013 Mousavi: Test Automation #12;Test Automation Outline Test Automation Mousavi: Test Automation #12;Test Automation Why? Challenges of Manual Testing Test-case design: Choosing inputs

Mousavi, Mohammad

340

GASIFICATION TEST RUN TC06  

SciTech Connect (OSTI)

This report discusses test campaign TC06 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC06. Test run TC06 was started on July 4, 2001, and completed on September 24, 2001, with an interruption in service between July 25, 2001, and August 19, 2001, due to a filter element failure in the PCD caused by abnormal operating conditions while tuning the main air compressor. The reactor temperature was varied between 1,725 and 1,825 F at pressures from 190 to 230 psig. In TC06, 1,214 hours of solid circulation and 1,025 hours of coal feed were attained with 797 hours of coal feed after the filter element failure. Both reactor and PCD operations were stable during the test run with a stable baseline pressure drop. Due to its length and stability, the TC06 test run provided valuable data necessary to analyze long-term reactor operations and to identify necessary modifications to improve equipment and process performance as well as progressing the goal of many thousands of hours of filter element exposure.

Southern Company Services, Inc.

2003-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

FUSION POWER Tokamak Test a Big Success  

Science Journals Connector (OSTI)

FUSION POWER Tokamak Test a Big Success ... Successful plasma production in the tokamak fusion test reactor at Princeton University's Plasma Physics Laboratory has set the stage for possible sustained fusion reactions in the device by 1990. ...

1983-01-10T23:59:59.000Z

342

Methods for quantifying uncertainty in fast reactor analyses.  

SciTech Connect (OSTI)

Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

Fanning, T. H.; Fischer, P. F.

2008-04-07T23:59:59.000Z

343

The thermal regime in the resurgent dome of Long Valley Caldera, California: Inferences from precision temperature logs in deep wells  

Science Journals Connector (OSTI)

Long Valley Caldera in eastern California formed 0.76 Ma ago in a cataclysmic eruption that resulted in the deposition of 600 km3 of Bishop Tuff. The total current heat flow from the caldera floor is estimated to be ~ 290 MW, and a geothermal power plant in Casa Diablo on the flanks of the resurgent dome (RD) generates ~40 MWe. The RD in the center of the caldera was uplifted by ~ 80 cm between 1980 and 1999 and was explained by most models as a response to magma intrusion into the shallow crust. This unrest has led to extensive research on geothermal resources and volcanic hazards in the caldera. Here we present results from precise, high-resolution, temperature–depth profiles in five deep boreholes (327–1,158 m) on the RD to assess its thermal state, and more specifically 1) to provide bounds on the advective heat transport as a guide for future geothermal exploration, 2) to provide constraints on the occurrence of magma at shallow crustal depths, and 3) to provide a baseline for future transient thermal phenomena in response to large earthquakes, volcanic activity, or geothermal production. The temperature profiles display substantial non-linearity within each profile and variability between the different profiles. All profiles display significant temperature reversals with depth and temperature gradients Valley boreholes are at the approximate same elevation as the high-temperature unit in borehole M-1 in Casa Diablo indicating lateral or sub-lateral hydrothermal flow through the resurgent dome. Small differences in temperature between measurements in consecutive years in three of the wells suggest slow cooling of the shallow hydrothermal flow system. By matching theoretical curves to segments of the measured temperature profiles, we calculate horizontal groundwater velocities in the hydrothermal flow unit under the RD that range from 1.9 to 2.8 m/yr, which corresponds to a maximum power flowing through the RD of 3–4 MW. The relatively low temperatures and large isothermal segments at the bottom of the temperature profiles are inconsistent with the presence of magma at shallow crustal levels.

Shaul Hurwitz; Christopher D. Farrar; Colin F. Williams

2010-01-01T23:59:59.000Z

344

B Reactor | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its

345

Development of Materials for Supercritical-Water-Cooled Reactor |  

Broader source: Energy.gov (indexed) [DOE]

Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Development of Materials for Supercritical-Water-Cooled Reactor Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system simplification, the R&D cost minimization and the flexibility for core design. As the demand for advanced nuclear system increases, Japanese R&D project started in 1999 aiming to provide technical information essential to demonstration of SCPR technologies through three sub-themes of 1. Plant conceptual design, 2. Thermal-hydraulics, and 3. Material. Although the material development is critical issue of SCWR development, previous studies were limited for the screening tests on commercial alloys

346

High Flux Isotope Reactor (HFIR) | Nuclear Science | ORNL  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

High Flux Isotope Reactor High Flux Isotope Reactor May 30, 2013 The High Flux Isotope Reactor (HFIR) first achieved criticality on August 25, 1965, and achieved full power in August 1966. It is a versatile 85-MW isotope production, research, and test reactor with the capability and facilities for performing a wide variety of irradiation experiments and a world-class neutron scattering science program. HFIR is a beryllium-reflected, light water-cooled and moderated flux-trap type swimming pool reactor that uses highly enriched uranium-235 as fuel. HFIR typically operates seven 23-to-27 day cycles per year. Irradiation facility capabilities include Flux trap positions: Peak thermal flux of 2.5X1015 n/cm2/s with similar epithermal and fast fluxes (Highest thermal flux available in the

347

EIS-0075: Strategic Petroleum Reserve Phase III Development, Texoma and Seaway Group Salt Domes (West Hackberry and Bryan Mound Expansion, Big Hill Development) Cameron Parish, Louisiana, and Brazoria and Jefferson Counties, Texas  

Broader source: Energy.gov [DOE]

Also see EIS-0021 and EIS-0029. The Strategic Petroleum Reserve (SPR) Office developed this EIS to assess the environmental impacts of expanding the existing SPR storage capacity from 538 million to 750 million barrels of storage and increasing the drawdown capability from 3.5 million to 4.5 million barrels per day. This EIS incorperates two previously issued EISs: DOE/EIS-0021, Seaway Group of Salt Domes, and DOE/EIS-0029, Texoma Group of Salt Domes.

348

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core  

ScienceCinema (OSTI)

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

None

2014-05-07T23:59:59.000Z

349

Stratigraphic and structural analysis of Shannon Sandstone, Teapot Dome field: implications for secondary and tertiary oil recovery  

SciTech Connect (OSTI)

The Department of Energy and Lawrence-Allison Associates have initiated three enhanced oil recovery (EOR) pilot projects in the last 4 years in the Shannon Sandstone at the Teapot Dome field. Performance of these pilot projects has generally been poor. As a result, a reevaluation of the geology for the entire field and for the pilot areas was conducted in an attempt to explain the pilots' performances. Based on core descriptions, conceptual reservoir flow patterns, and post-fireflood coring, only the bar-margin facies was found amenable to fluid displacement processes for oil recovery. This results in 25 million bbl of oil originally in place vs 180 million bbl of oil originally in place as a target for EOR. Stratigraphy alone does not explain the observed production patterns and simulation of the reservoir. Faulting and fracturing are extensive. Based on faults mapped in the in-situ pilot area, 12 to 16 northeast-trending normal faults per mile can be projected. Fracture orientations were obtained by mapping calcareous streaks in the upper and lower Shannon. These orientations confirm directions of premature fluid breakthroughs observed in the pilot projects and in the old East Teapot waterflood. Water resistivities and total dissolved solids measurements, water-cut maps, and daily oil production maps suggest that some faults are partial to total barriers to fluid flow across the field.

Chappelle, H.; Emsurak, G.; Obernyer, S.

1986-08-01T23:59:59.000Z

350

LLNL-Generated Content for the California Academy of Sciences, Morrison Planetarium Full-Dome Show: Earthquake  

SciTech Connect (OSTI)

The California Academy of Sciences (CAS) Morrison Planetarium is producing a 'full-dome' planetarium show on earthquakes and asked LLNL to produce content for the show. Specifically the show features numerical ground motion simulations of the M 7.9 1906 San Francisco and a possible future M 7.05 Hayward fault scenario earthquake. The show also features concepts of plate tectonics and mantle convection using images from LLNL's G3D global seismic tomography. This document describes the data that was provided to the CAS in support of production of the 'Earthquake' show. The CAS is located in Golden Gate Park, San Francisco and hosts over 1.6 million visitors. The Morrison Planetarium, within the CAS, is the largest all digital planetarium in the world. It features a 75-foot diameter spherical section projection screen tilted at a 30-degree angle. Six projectors cover the entire field of view and give a three-dimensional immersive experience. CAS shows strive to use scientifically accurate digital data in their productions. The show, entitled simply 'Earthquake', will debut on 26 May 2012. They are working on graphics and animations based on the same data sets for display on LLNL powerwalls and flat-screens as well as for public release.

Rodgers, A J; Petersson, N A; Morency, C E; Simmons, N A; Sjogreen, B

2012-01-23T23:59:59.000Z

351

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

352

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

353

Reactor for exothermic reactions  

DOE Patents [OSTI]

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02T23:59:59.000Z

354

Diagnostics for hybrid reactors  

SciTech Connect (OSTI)

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19T23:59:59.000Z

355

Evaluation on the Feasibility of Using Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density/Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock  

SciTech Connect (OSTI)

This technical letter report provides the status of an assessment undertaken by PNNL at the request of the NRC to verify the capability of periodic ASME-required volumetric examinations of reactor vessels to characterize the density and distribution of flaws of interest for applying §50.61a on a plant-by-plant basis. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. Recently, the NRC completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed and promulgated an alternate PTS rule, §50.61a, that can be implemented by PWR licensees. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants.

Sullivan, Edmund J.; Anderson, Michael T.

2014-06-10T23:59:59.000Z

356

Distribution of nanoscale nuclei in the amorphous dome of a phase change random access memory  

SciTech Connect (OSTI)

The nanoscale crystal nuclei in an amorphous Ge{sub 2}Sb{sub 2}Te{sub 5} bit in a phase change memory device were evaluated by fluctuation transmission electron microscopy. The quench time in the device (?10 ns) afforded more and larger nuclei in the melt-quenched state than in the as-deposited state. However, nuclei were even more numerous and larger in a test structure with a longer quench time (?100 ns), verifying the prediction of nucleation theory that slower cooling produces more nuclei. It also demonstrates that the thermal design of devices will strongly influence the population of nuclei, and thus the speed and data retention characteristics.

Lee, Bong-Sub, E-mail: bongsub@gmail.com; Darmawikarta, Kristof; Abelson, John R. [Department of Materials Science and Engineering and the Coordinated Sciences Laboratory, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States); Raoux, Simone; Shih, Yen-Hao; Zhu, Yu [IBM/Macronix PCRAM Joint Project, IBM T. J. Watson Research Center, Yorktown Heights, New York 10598 (United States); Bishop, Stephen G. [Department of Materials Science and Engineering and the Coordinated Sciences Laboratory, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States); Department of Electrical and Computer Engineering and the Coordinated Sciences Laboratory, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States)

2014-02-17T23:59:59.000Z

357

Structural materials for fusion reactors  

Science Journals Connector (OSTI)

Fusion Reactors will require specially engineered structural materials, which ... on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics ...

P. M. Raole; S. P. Deshpande

2009-04-01T23:59:59.000Z

358

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect (OSTI)

Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

359

Reactor Materials | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Benefits Crosscutting Technology Development Reactor Materials Advanced Sensors and Instrumentation Proliferation and Terrorism Risk Assessment Advanced Methods for Manufacturing...

360

Reactor operation safety information document  

SciTech Connect (OSTI)

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Not Available

1990-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

Fossil fuel furnace reactor  

DOE Patents [OSTI]

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

362

Thermal Reactor Safety  

SciTech Connect (OSTI)

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Not Available

1980-06-01T23:59:59.000Z

363

Reactor Safety Research Programs Quarterly Report October - December 1981  

SciTech Connect (OSTI)

This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1982-03-01T23:59:59.000Z

364

Bibliography of reports by US Geological Survey personnel pertaining to underground nuclear testing and radioactive waste disposal at the Nevada Test Site, and radioactive waste disposal at the WIPP Site, New Mexico, January 1, 1979-December 31, 1979  

SciTech Connect (OSTI)

This bibliography presents reports released to the public between January 1, 1979, and December 31, 1979, by personnel of the US Geological Survey. Reports include information on underground nuclear testing and waste management projects at the NTS (Nevada Test Site) and radioactive waste projects at the WIPP (Waste Isolation Pilot Plant) site, New Mexico. Reports on Project Dribble, Tatum Dome, Mississippi, previously prepared as administrative reports and released to the public as 474-series reports during 1979 are also included in this bibliography.

Glanzman, V.M.

1980-01-01T23:59:59.000Z

365

Radiation effects on reactor pressure vessel supports  

SciTech Connect (OSTI)

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

1996-05-01T23:59:59.000Z

366

Reactor vessel support system. [LMFBR  

DOE Patents [OSTI]

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Golden, M.P.; Holley, J.C.

1980-05-09T23:59:59.000Z

367

SciTech Connect: ABRASION TEST OF THORIA PELLETS  

Office of Scientific and Technical Information (OSTI)

BREEDING; PELLETS; REACTORS; TESTING; THERMAL NEUTRONS; THORIUM; THORIUM OXIDES Word Cloud More Like This Full Text preview image File size NAView Full Text View Full Text DOI:...

368

Fusion reactor systems  

Science Journals Connector (OSTI)

In this review we consider deuterium-tritium (D-T) fusion reactors based on four different plasma-confinement and heating approaches: the tokamak, the theta-pinch, the magnetic-mirror, and the laser-pellet system. We begin with a discussion of the dynamics of reacting plasmas and basic considerations of reactor power balance. The essential plasma physical aspects of each system are summarized, and the main characteristics of the corresponding conceptual power plants are described. In tokamak reactors the plasma densities are about 1020 m-3, and the ? values (ratio of plasma pressure to confining magnetic pressure) are approximately 5%. Plasma burning times are of the order of 100-1000 sec. Large superconducting dc magnets furnish the toroidal magnetic field, and 2-m thick blankets and shields prevent heat deposition in the superconductor. Radially diffusing plasma is diverted away from the first wall by means of null singularities in the poloidal (or transverse) component of the confining magnetic field. The toroidal theta-pinch reactor has a much smaller minor diameter and a much larger major diameter, and operates on a 10-sec cycle with 0.1-sec burning pulses. It utilizes shock heating from high-voltage sources and adabatic-compression heating powered by low-voltage, pulsed cryogenic magnetic or inertial energy stores, outside the reactor core. The plasma has a density of about 1022 m-3 and ? values of nearly unity. In the power balance of the reactor, direct-conversion energy obtained by expansion of the burning high-? plasma against the containing magnetic field is an important factor. No divertor is necessary since neutral-gas flow cools and replaces the "spent" plasma between pulses. The open-ended mirror reactor uses both thermal conversion of neutron energy and direct conversion of end-loss plasma energy to dc electrical power. A fraction of this direct-convertor power is then fed back to the ioninjection system to sustain the reaction and maintain the plasma. The average ion energy is 600 keV, plasma diameter 6 m, and the plasma beta 85%. The power levels of the three magnetic-confinement devices are in the 500-2000 MWe range, with the exception of the mirror reactor, for which the output is approximately 200 MWe. In Laser-Pellet reactors, frozen D-T pellets are ignited in a cavity which absorbs the electromagnetic, charged particle, and neutron energy from the fusion reaction. The confinement is "inertial," since the fusion reaction occurs during the disassembly of the heated pellet. A pellet-cavity unit would produce about 200 MWt in pulses with a repetition rate of the order of 10 sec-1. Such units could be clustered to give power plants with outputs in the range of 1000 MWe.

F. L. Ribe

1975-01-01T23:59:59.000Z

369

Nuclear reactor construction with bottom supported reactor vessel  

DOE Patents [OSTI]

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

1987-01-01T23:59:59.000Z

370

Spherical torus fusion reactor  

DOE Patents [OSTI]

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

371

Nuclear divisional reactor  

SciTech Connect (OSTI)

A nuclear divisional reactor including a reactor core having side and top walls, a heat exchanger substantially surrounding the core, the heat exchanger including a plurality of separate fluid holding and circulating chambers each in contact with a portion of the core, control rod means associated with the core and external of the heat exchanger including control rods and means for moving said control rods, each of the chambers having separate means for delivering and removing fluid therefrom, separate means associated with each of the delivering and removing means for producing useable energy external of the chambers, each of the means for producing useable energy having separate variable capacity energy outputs thereby making available a plurality of individual sources of useable energy of varying degrees.

Administratrix, A.P.; Rugh, J.L.

1982-11-02T23:59:59.000Z

372

Fusion reactor pumped laser  

DOE Patents [OSTI]

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Jassby, D.L.

1987-09-04T23:59:59.000Z

373

Task summary for cone penetrating testing sounding and soil and groundwater sampling Salmon Site, Lamar County, Mississippi  

SciTech Connect (OSTI)

The Salmon Site (SS), located in Mississippi, was the site of two nuclear and two gas explosion testes conducted deep underground in the Tatum Salt Dome between 1964 and 1970. As a consequence radionuclides generated during the testing were released into the salt dome. During reentry drilling and other site activities, incidental liquid and solid wastes that contained radioactivity were generated, resulting in some soil, ground water and equipment contamination. US DOE is conducting a series of investigations as a part of the Remedial Investigation and Feasibility Study (under CERCLA) This report summarizes the cone penetrometer testing (CPT) and sampling program conducted in fall 1993, providing a description of the activities and a discussion of the results. The objectives of the CPT program were to determine subsurface conditions and stratification; determine the depth to the potentiometric surface; obtain soil samples from predetermined depths; obtain groundwater samples at predetermined depths.

Not Available

1994-10-01T23:59:59.000Z

374

Solution of the space-time reactor kinetics equations using the method of Laplace transforms  

E-Print Network [OSTI]

ANALYTIC CALCULATIONS The analytic solut1on developed in Chapter II is the basis for the computer program ANALYTIC, which calculates the time and spatially dependent solution for the neutron flux and the precursor concentration in a bare slab reactor... case to test the full capabilities of the ANALYTIC computer program is the subcrit1cal reactor. The reactor parameters for the thermal, subcritical and the fast, subcrit1cal reacto~ is listed 1n Table IV. The solution for the neutron flux...

Rottler, Jerry Stephen

2012-06-07T23:59:59.000Z

375

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database  

SciTech Connect (OSTI)

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

Wang, Jy-An John [ORNL

2010-08-01T23:59:59.000Z

376

Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods  

SciTech Connect (OSTI)

Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs.

Rothrock, R.B.

1991-01-01T23:59:59.000Z

377

Mirror Advanced Reactor Study (MARS) final report summary  

SciTech Connect (OSTI)

The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory.

Henning, C.D.; Logan, B.G.; Carlson, G.A.; Perkins, L.J.; Werner, R.W.; Gordon, J.D.; Parmer, J.F.; Bilton, J.R.; Glancy, J.E.; Kulcinski, G.L.

1983-12-08T23:59:59.000Z

378

Gas Test Loop Facilities Alternatives Assessment Report Rev 1  

SciTech Connect (OSTI)

An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL.

William J. Skerjanc; William F. Skerjanc

2005-07-01T23:59:59.000Z

379

Nuclear Reactor Materials and Fuels  

Science Journals Connector (OSTI)

Nuclear reactor materials and fuels can be classified into six categories: Nuclear fuel materials Nuclear clad materials Nuclear coolant materials Nuclear poison materials Nuclear moderator materials

Dr. James S. Tulenko

2012-01-01T23:59:59.000Z

380

Thermonuclear Reflect AB-Reactor  

E-Print Network [OSTI]

The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

Alexander Bolonkin

2008-03-26T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Light Water Reactor Sustainability Newsletter  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

hydraulics software RELAP-7 (which is under development in the Light Water Reactor Sustainability LWRS Program). A novel interaction between the probabilistic part (i.e., RAVEN)...

382

Savannah River Site | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Savannah River Site Savannah River Site Savannah River Site Work is under way to decommission the Heavy Water Components Test Reactor, which had been used to test experimental fuel assemblies for commercial heavy-water power reactors. SRS is scheduled to remove the dome of the reactor this month (January 2011). Workers also will displace the reactor vessel and steam generators, grout the remaining structure in place, and install a concrete cover over the reactor's footprint Work is under way to decommission the Heavy Water Components Test Reactor, which had been used to test experimental fuel assemblies for commercial heavy-water power reactors. SRS is scheduled to remove the dome of the reactor this month (January 2011). Workers also will displace the reactor vessel and steam generators, grout the remaining structure in place, and

383

Reactor coolant pump flywheel  

DOE Patents [OSTI]

A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

2013-11-26T23:59:59.000Z

384

Biparticle fluidized bed reactor  

DOE Patents [OSTI]

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

Scott, C.D.; Marasco, J.A.

1995-04-25T23:59:59.000Z

385

Biparticle fluidized bed reactor  

DOE Patents [OSTI]

A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

Scott, C.D.; Marasco, J.A.

1996-02-27T23:59:59.000Z

386

Nuclear reactor control apparatus  

DOE Patents [OSTI]

Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

Sridhar, Bettadapur N. (Cupertino, CA)

1983-11-01T23:59:59.000Z

387

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power  

SciTech Connect (OSTI)

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.

John Darrell Bess

2008-06-01T23:59:59.000Z

388

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network [OSTI]

Modular Pebble Bed Reactor High Temperature Gas Reactor Andrew C Kadak Massachusetts Institute For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR/Graphite Discrimination system Damaged Sphere ContainerGraphiteReturn FuelReturn Fresh Fuel Container Spent Fuel Tank #12

389

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

390

Piezoelectric material for use in a nuclear reactor core  

SciTech Connect (OSTI)

In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

Parks, D. A.; Reinhardt, Brian; Tittmann, B. R. [EES Department, Penn State University, University Park, PA 16802 (United States)

2012-05-17T23:59:59.000Z

391

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow  

SciTech Connect (OSTI)

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15T23:59:59.000Z

392

DOE Drops Plan to Restart Reactor  

Science Journals Connector (OSTI)

...longer in flux. Hanford research reactor...decision to scrap the Hanford reactor, which...research. At public meetings, however...decision to scrap the Hanford reactor, which...research. At public meetings, however, FFTF...

Robert F. Service

2000-12-01T23:59:59.000Z

393

Operational Analysis of Multiregional Nuclear Reactor Kinetics  

Science Journals Connector (OSTI)

......Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAR H. S. HAIDAR...analytically for a multiregional nuclear reactor whose subregions are of arbitrary...Operational Analysis of Multiregional Nuclear Reactor Kinetics NASSAU H. S. HAIDAR......

NASSAR H. S. HAIDAR

1983-05-01T23:59:59.000Z

394

Solvent refined coal reactor quench system  

DOE Patents [OSTI]

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Thorogood, R.M.

1983-11-08T23:59:59.000Z

395

Partnerships Help Advance Small Modular Reactor Technology | Department of  

Broader source: Energy.gov (indexed) [DOE]

Partnerships Help Advance Small Modular Reactor Technology Partnerships Help Advance Small Modular Reactor Technology Partnerships Help Advance Small Modular Reactor Technology March 5, 2012 - 12:00pm Addthis WASHINGTON, D.C. - DOE recently announced three public-private partnerships to develop deployment plans for small modular nuclear reactor (SMR) technologies at Savannah River Site (SRS) facilities near Aiken, S.C. Read the full story on the Memorandums of Agreement to help leverage SRS land assets, energy facilities and nuclear expertise to support potential private sector development, testing and licensing of prototype SMR technologies. Addthis Related Articles Energy Department Announces Small Modular Reactor Technology Partnerships at Savannah River Site The development of clean, affordable nuclear power options is a key element of the Energy Department's Nuclear Energy Research and Development Roadmap. As a part of this strategy, a high priority of the Department has been to help accelerate the timelines for the commercialization and deployment of small modular reactor (SMR) technologies through the SMR Licensing Technical Support program. | Photo by the Energy Department.

396

Temperature effects on chemical reactor  

Science Journals Connector (OSTI)

In this paper we had to study some characteristics of the chemical reactors from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature it is a mathematical processing of a chemical problem that was already studied but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid?gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology especially in our century.

M. Azzouzi

2008-01-01T23:59:59.000Z

397

THE MATERIALS OF FAST BREEDER REACTORS  

E-Print Network [OSTI]

metal fast breeder reactor (LMFBR) concern the behavior ofmetal fast breeder reactor (LMFBR). Despite the simplicityinduced by irradiation. LMFBR funding is the largest single

Olander, Donald R.

2013-01-01T23:59:59.000Z

398

Nuclear reactors in the United States  

Science Journals Connector (OSTI)

Nuclear reactors in the United States ... A chart listing the operating and planned nuclear reactors in the United States. ... Nuclear / Radiochemistry ...

Hubert N. Alyea

1956-01-01T23:59:59.000Z

399

Advanced Reactor Research and Development Funding Opportunity...  

Broader source: Energy.gov (indexed) [DOE]

Reactor Research and Development Funding Opportunity Announcement Advanced Reactor Research and Development Funding Opportunity Announcement The U.S. Department of Energy (DOE)...

400

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

LIGHT WATER REACTOR SUSTAINABILITY PROGRAM: INTRODUCTION The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1...

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

MOOSE simulating nuclear reactor CRUD buildup  

SciTech Connect (OSTI)

This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

None

2014-02-06T23:59:59.000Z

402

Advanced Reactor Technologies | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

403

Granular Dynamics in Pebble Bed Reactor Cores  

E-Print Network [OSTI]

pebble bed reactor,” Nuclear Engineering and Design, vol.the AVR reactor,” Nuclear Engineering and Design, vol. 121,Operating Experience,” Nuclear Engineering and Design, vol.

Laufer, Michael Robert

2013-01-01T23:59:59.000Z

404

F Reactor Inspection | Department of Energy  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Inspection F Reactor Inspection Addthis Description Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor last week before...

405

Physics of nuclear reactor safety  

Science Journals Connector (OSTI)

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive or natural safety. The second chapter focuses on the basics of reactor safety, identifying the main risk sources and the main principles for a safe design. The third chapter concerns a systematic treatment of the physical processes that are fundamental for the properties of fission chain reacting processes and the control of those processes. Because of the rather specialized nature of the field of reactor physics, each paragraph contains a very concise description of the theory of the phenomenon under consideration, before presenting a review of the developments. Chapter 4 contains a short review of the thermal aspects of reactor safety, restricted to those aspects that are characteristic of the nuclear reactor field, because thermal hydraulics of fission reactors is not principally different from that of other physical systems. In chapter 5 the consequences of the physics treated in the preceding chapters for the dynamics and safety of actual reactors are reviewed. The systematics of the treatment is mainly based on a division of reactors into three categories according to the type of coolant, which to a large extent determines the safety properties of the reactors. The last chapter contains a physical analysis of the Chernobyl accident that occurred in 1986. The reason for an attempt to give a review of this accident, as complete as possible within the space limits set by the editors, is twofold: the Chernobyl accident is the most severe accident in history and physical properties of the reactor played a decisive role, thereby serving as an illustration of the material of the preceding chapters.

H van Dam

1992-01-01T23:59:59.000Z

406

Process characterization of expanded-bed reactors in series  

Science Journals Connector (OSTI)

A process model describing the performance of a resid hydrotreater has been developed using lumped first-order kinetic equations solved for three back-mixed reactors in series. The model has been tested against pilot-plant and commercial data and found to give a good representation of the measurements.

J.F. Mosby; R.D. Buttke; J.A. Cox; C. Nikolaides

1986-01-01T23:59:59.000Z

407

Development of advanced strain diagnostic techniques for reactor environments.  

SciTech Connect (OSTI)

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

2013-02-01T23:59:59.000Z

408

13 - Generation IV reactor designs, operation and fuel cycle  

Science Journals Connector (OSTI)

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

N. Cerullo; G. Lomonaco

2012-01-01T23:59:59.000Z

409

Spectral Effects on Stress Relaxation of Inconel X-750 Springs in CANDU Reactors  

SciTech Connect (OSTI)

CANDU reactors have been operating for periods up to about 25 years. During this time there are changes to the nuclear reactor core components that are a function of operating environment and time. It is important to know how the properties of critical core components are likely to change over the life of a reactor and therefore their behaviours are characterised long before the end of the reactor design life. Tests are typically conducted in materials test reactors. The behaviour of a material is often characterised as a function of fast neutron fluence and the expected effect of operating time is established by simply extrapolating as a function of fluence. This may be appropriate when the neutron energy spectrum for the materials test reactor matches closely the neutron spectrum where the component resides in the power reactor. However, in cases where the spectrum is very different one has to convert the accumulated dose into a unit that is common in its effect on the material properties. For many property changes in nuclear reactor cores this unit is displacements per atom (dpa).

Griffiths, M.; Butcher, F. J.; Ariani, I.; Douglas, S.; Garner, Francis A.; Greenwood, Lawrence R.

2008-11-16T23:59:59.000Z

410

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor  

SciTech Connect (OSTI)

The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

Eric C. Woolstenhulme; Dana M. Hewit

2008-09-01T23:59:59.000Z

411

In-reactor oxidation of zircaloy-4 under low water vapor pressures  

SciTech Connect (OSTI)

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Walter G. Luscher; David J. Senor; Keven K. Clayton; Glen R. Longhurst

2015-01-01T23:59:59.000Z

412

Modeling a Catalytic Reactor for Hydrotreating of Straight-Run Gas Oil Blended with Fluid Catalytic Cracking Naphtha and Light Cycle Oil: Influence of Vapor–Liquid Equilibrium  

Science Journals Connector (OSTI)

Model results were validated using the industrial test run data, and very good predictions of overall sulfur conversion and reactor temperature were obtained. ...

Ivana M. Mijatovi?; Sandra B. Glisic; Aleksandar M. Orlovi?

2014-11-17T23:59:59.000Z

413

Expert system for online surveillance of nuclear reactor coolant pumps  

DOE Patents [OSTI]

An expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

Gross, Kenny C. (Bolingbrook, IL); Singer, Ralph M. (Naperville, IL); Humenik, Keith E. (Columbia, MD)

1993-01-01T23:59:59.000Z

414

Assessment of torsatrons as reactors  

SciTech Connect (OSTI)

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. (Oak Ridge National Lab., TN (United States)) [Oak Ridge National Lab., TN (United States); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia)) [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

415

Assessment of torsatrons as reactors  

SciTech Connect (OSTI)

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R{sub 0} = 6.6-8.8 m, on-axis magnetic field B{sup 0} = 4.8-7.5 T, B{sub max} (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Painter, S.L. [Australian National Univ., Canberra, ACT (Australia)

1992-12-01T23:59:59.000Z

416

SDI: Solar Dome Instrument for Solar Irradiance Monitoring Tao Liu1, Ankur U. Kamthe1, Varick L. Erickson1, Carlos F. M. Coimbra2 and Alberto E. Cerpa1  

E-Print Network [OSTI]

SDI: Solar Dome Instrument for Solar Irradiance Monitoring Tao Liu1, Ankur U. Kamthe1, Varick L data for ground solar irradiance (direct normal and global irradiance) is a major obstacle for the de- velopment of adequate policies to promote and take advan- tage of existing solar technologies. Although

Cerpa, Alberto E.

417

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Broader source: Energy.gov (indexed) [DOE]

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

418

Nuclear reactor downcomer flow deflector  

DOE Patents [OSTI]

A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

2011-02-15T23:59:59.000Z

419

Testability of a heat pipe cooled thermionic reactor  

Science Journals Connector (OSTI)

As part of the Air Force Phillips Laboratory thermionics program Rocketdyne performed a design study for an in?core thermionic fuel element (TFE) heat pipe cooled reactor power system. This effort involved a testability evaluation that was performed starting with testing of individual components followed by testing at various stages of fabrication and concluding with full system acceptance and qualification testing. It was determined that the system could be thoroughly tested to ensure a high probability of successful operation in space after launch.

Richard E. Durand

1992-01-01T23:59:59.000Z

420

Integrated intelligent systems in advanced reactor control rooms  

SciTech Connect (OSTI)

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

Beckmeyer, R.R.

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

2012 Annual Report Research Reactor Infrastructure Program  

SciTech Connect (OSTI)

The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

Douglas Morrell

2012-11-01T23:59:59.000Z

422

Tritium diagnostics in a fusion reactor  

Science Journals Connector (OSTI)

Methods for controlling tritium in a fusion reactor are reviewed. The characteristic features of the...

A. I. Markin; N. I. Syromyatnikov; A. M. Belov

2010-05-01T23:59:59.000Z

423

COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY  

SciTech Connect (OSTI)

This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

Vierow, Karen

2008-09-26T23:59:59.000Z

424

Black-box System Testing of Real-Time Embedded Systems Using Random and Search-based Testing  

E-Print Network [OSTI]

of nuclear reactors and flying systems). Testing RTES is particularly chal- lenging since they operateBlack-box System Testing of Real-Time Embedded Systems Using Random and Search-based Testing Andrea. Testing real-time embedded systems (RTES) is in many ways chal- lenging. Thousands of test cases can

Boyer, Edmond

425

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation  

SciTech Connect (OSTI)

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

2005-09-27T23:59:59.000Z

426

Combustion synthesis continuous flow reactor  

DOE Patents [OSTI]

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

1998-01-01T23:59:59.000Z

427

Interfacial effects in fast reactors  

E-Print Network [OSTI]

The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

Saidi, Mohammad Said

1979-01-01T23:59:59.000Z

428

Unique features of space reactors  

SciTech Connect (OSTI)

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

Buden, D.

1990-01-01T23:59:59.000Z

429

Nuclear Reactors and Technology; (USA)  

SciTech Connect (OSTI)

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01T23:59:59.000Z

430

Reactor physics project final report  

E-Print Network [OSTI]

This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

Driscoll, Michael J.

1970-01-01T23:59:59.000Z

431

Alternate-fuel reactor studies  

SciTech Connect (OSTI)

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01T23:59:59.000Z

432

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility  

SciTech Connect (OSTI)

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test.

Lee, Sukho [Korea Institute of Nuclear Safety (Korea, Republic of); Kim, Manwoong [Korea Institute of Nuclear Safety (Korea, Republic of); Kim, Hho-Jung [Korea Institute of Nuclear Safety (Korea, Republic of); Lee, John C. [University of Michigan (United States)

2005-09-15T23:59:59.000Z

433

Solar solids reactor  

DOE Patents [OSTI]

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, Bernard D. (Chicago, IL)

1987-01-01T23:59:59.000Z

434

Solar solids reactor  

DOE Patents [OSTI]

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, B.D.

1986-02-24T23:59:59.000Z

435

Novel Catalytic Membrane Reactors  

SciTech Connect (OSTI)

There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

Stuart Nemser, PhD

2010-10-01T23:59:59.000Z

436

Evaluation of Torsatrons as reactors  

SciTech Connect (OSTI)

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1994-03-01T23:59:59.000Z

437

When Do Commercial Reactors Permanently Shut Down?  

Reports and Publications (EIA)

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

2011-01-01T23:59:59.000Z

438

Fabrication and Characterization of Uranium-based High Temperature Reactor  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Fabrication and Characterization of Uranium-based High Temperature Reactor Fabrication and Characterization of Uranium-based High Temperature Reactor Fuel June 01, 2013 The Uranium Fuel Development Laboratory is a modern R&D scale lab for the fabrication and characterization of uranium-based high temperature reactor fuel. A laboratory-scale coater manufactures tri-isotropic (TRISO) coated fuel particles (CFPs), state-of-the-art materials property characterization is performed, and the CFPs are then pressed into fuel compacts for irradiation testing, all under a NQA-1 compliant Quality Assurance Program. After fuel kernel size and shape are measured by optical shadow imaging, the TRISO coatings are deposited via fluidized bed chemical vapor deposition in a 50-mm diameter conical chamber within the coating furnace. Computer control of temperature and gas composition ensures reproducibility

439

NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE  

SciTech Connect (OSTI)

The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess

2014-03-01T23:59:59.000Z

440

Safety philosophy of gas turbine high temperature reactor (GTHTR300)  

SciTech Connect (OSTI)

Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa [Department of Advanced Nuclear Heat Technology, Oarai Research Institute, Japan Atomic Energy Research Institute, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

2002-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor dome" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


441

Transpiring wall supercritical water oxidation reactor salt deposition studies  

SciTech Connect (OSTI)

Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G. [and others

1996-09-01T23:59:59.000Z

442

Code validation with EBR-II test data  

SciTech Connect (OSTI)

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experimental Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-II, and are also valuable tools for the analysis of innovative reactor designs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1992-01-01T23:59:59.000Z

443

Code validation with EBR-II test data  

SciTech Connect (OSTI)

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experimental Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-II, and are also valuable tools for the analysis of innovative reactor designs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1992-07-01T23:59:59.000Z

444

Code validation with EBR-II test data  

SciTech Connect (OSTI)

An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experiment Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-2, and are also valuable tools for the analysis of innovative reactor designs. 29 refs., 6 figs.

Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D.; Planchon, H.P.

1991-01-01T23:59:59.000Z

445

Microsoft Word - 911133_0 SSC-1 Test Plan_rel.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for Reactor Control Equipment Prepared by General Atomics For the Battelle Energy...

446

Microsoft Word - 911141_0_SSC-9 RCCS Test Plan.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan for the Reactor Cavity Cooling System Prepared by General Atomics For the Battelle...

447

Microsoft Word - 911137_0 SSC-5 RPV Test Plan_rel.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

SERVICES FOR THE NEXT GENERATION NUCLEAR PLANT (NGNP) WITH HYDROGEN PRODUCTION Test Plan - Reactor Pressure Vessel Prepared by General Atomics For the Battelle Energy...

448

Strategic Petroleum Reserve (SPR) oil storage cavern sulfur mines 7. Certification tests and analysis  

SciTech Connect (OSTI)

Cavern 7 at the Sulphur Mines, Louisiana SPR oil storage site was certified for oil storage on December 17, 1977. The Dowell Sonar caliper survey taken November 29, 1977, indicated a total cavern volume of 5.60 x 10/sup 6/ bbls. The surveys taken December 19, 1979, and June 10, 1981, indicated a total cavern volume of 6.33 x 10/sup 6/ and 6.36 x 10/sup 6/ bbls respectively. This volume increase was a result of continued brining, prior to June 10, 1981, to get brine enrichment for PPG. A well leak test in May 1981 indicated some well leakage. Well workover actions to repair well and wellhead leaks were taken by Texas Brine Corp/Dravo Utility Constructors, Inc. (TBC/DUCI). Testing was restarted in June 1981 using test procedures which were developed in conjunction with the procedures and testing of West Hackberry cavern 6. This report includes a general history of the cavern and a description of the certification testing, analyses, conclusions, and recommendations. The data from cavern 7 and 6 indicate no fluid communication between caverns. Cavern 7 is about 160 ft from the dome edge. The pressure data at maximum operating pressure is comparable to the data from both West Hackberry cavern 6 and Sulphur Mines cavern 6. Therefore, it is considered unlikely that there is a leak to the dome edge. The well test data indicates leaks in the well casing seat area are approximately 100 bbls/yr.

Beasley, R.R.

1982-05-01T23:59:59.000Z

449

Advanced Nuclear Reactors | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

450

Advanced Reactor Technology Documents | Department of Energy  

Broader source: Energy.gov (indexed) [DOE]

Nuclear Reactor Technologies » Advanced Reactor Nuclear Reactor Technologies » Advanced Reactor Technologies » Advanced Reactor Technology Documents Advanced Reactor Technology Documents January 30, 2013 Advanced Reactor Concepts Technical Review Panel Report This report documents the establishment of a technical review process and the findings of the Advanced Reactor Concepts (ARC) Technical Review Panel (TRP).1 The intent of the process is to identify R&D needs for viable advanced reactor concepts in order to inform DOE-NE R&D investment decisions. A goal of the process is to facilitate greater engagement between DOE and industry. The process involved establishing evaluation criteria, conducting a pilot review, soliciting concept inputs from industry entities, reviewing the concepts by TRP members and compiling the

451

Microsoft Word - power_reactors_briggs.doc  

Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

Most common - Boiling Water and Pressurized Most common - Boiling Water and Pressurized Water Reactors About 80% of the world's nuclear reactors used for generating electricity are either boiling water reactors or pressurized water reactors. Of these, about 30% are boiling water reactors and 70% are pressurized water reactors. All power reactors currently in use in the United States are of these two types. Both types of reactors have been very successfully used for reliable, on-demand, emissions-free electricity generation for decades. How does a boiling water reactor work? Water flows from the bottom of the fuel to the top of the fuel, and a